Targetry coupled separations

ABSTRACT

Targetry coupled separation refers to enhancing the production of a predetermined radiation product through the selection of a target (including selection of the target material and the material&#39;s physical structure) and separation chemistry in order to optimize the recovery of the predetermined radiation product. This disclosure describes systems and methods for creating (through irradiation) and removing one or more desired radioisotopes from a target and further describes systems and methods that allow the same target to undergo multiple irradiations and separation operations without damage to the target. In contrast with the prior art that requires complete dissolution or destruction of a target before recovery of any irradiation products, the repeated reuse of the same physical target allowed by targetry coupled separation represents a significant increase in efficiency and decrease in cost over the prior art.

RELATED APPLICATIONS

This application claims the benefit of U.S. Provisional Application No.62/097,235, filed Dec. 29, 2014, which application is herebyincorporated by reference in its entirety.

INTRODUCTION

Irradiation of chemical compounds and separating materials from theirradiated compound have a range of technical applications whichincludes the production of radioisotopes, nuclear fuel processing, andfundamental scientific research. For example, the following is a tablethat lists some valuable radioisotopes including those with medicalapplications. Some of the radioisotopes are generated directly fromprecursor fissionable material while others are decay products of otherradioisotopes.

TABLE 1 Radioisotopes and Uses Radioisotope Symbol Half Life Use(s)Actinium-227 ²²⁷Ac 21.8 y As a parent of ²²³Ra, used to create a ²²³Ragenerator (such an isotope generator may also sometimes be referred toas a ²²⁷Ac “cow” that is occasionally “milked” to obtain the ²²³Raisotope). Bismuth-213 ²¹³Bi 46 min Used for targeted alpha therapy(TAT), especially cancers, as it has a high energy (8.4 MeV). Cesium -various ^(xx)Cs Used in brachytherapy, particularly ¹³³Cs and isotopes¹³¹Cs. Carbon-11 ¹¹C 20.3 m Positron emitter used in positron emissiontomography (PET) for studying brain physiology and pathology, inparticular for localizing epileptic focus, and in dementia, psychiatryand neuropharmacology studies. Also has a role in cardiology.Chromium-51 ⁵¹Cr 27.7 d Used to label red blood cells and quantifygastro-intestinal protein loss. Cobalt-57 ⁵⁷Co 271.8 d Used as a markerto estimate organ size and for in-vitro diagnostic kits. Cobalt-60 ⁶⁰Co5.271 y Formerly used for external beam radiotherapy. Copper-64 ⁶⁴Cu12.7 h Used to study genetic diseases affecting copper metabolism, suchas Wilson's and Menke's diseases, and for PET imaging of tumors, andtherapy. Copper-67 ⁶⁷Cu 2.6 d Beta emitter, used in therapy.Dysprosium-165 ¹⁶⁵Dy 2.33 h Used as an aggregated hydroxide forsynovectomy treatment of arthritis. Erbium-169 ¹⁶⁹Er 9.4 d Used forrelieving arthritis pain in synovial joints. Fluorine-18 ¹⁸F 1.83 hPositron emitter used in PET for studying brain physiology andpathology, in particular for localizing epileptic focus, and indementia, psychiatry and neuropharmacology studies. Also has a role incardiology. ¹⁸F has become very important in detection of cancers andthe monitoring of progress in their treatment, using PET. Gallium-67⁶⁷Ga 78 h Used for tumor imaging and localization of inflammatorylesions (infections). Gallium-68 ⁶⁸Ga 68 min Positron emitter used inPET and PET-CT units. A daughter of ⁶⁸Ge typically obtained from a 68Gegenerator. Germanium-68 ⁶⁸Ge 271 d Used as the ‘parent’ in a generatorto produce the daughter isotope ⁶⁸Ga. Gold, various ^(xx)Au Used inbrachytherapy. ¹⁹⁸Au in particular is isotopes used for treatment of theprostate. Holmium-166 ¹⁶⁶Ho 26 h Being developed for diagnosis andtreatment of liver tumors. Indium-111 ¹¹¹In 2.8 d Used for specializeddiagnostic studies, e.g. brain studies, infection and colon transitstudies. Iodine-123 ¹²³I 13.2 h Increasingly used for diagnosis ofthyroid function, it is a gamma emitter from electron capture. Thisisotope does not have the beta decay that occurs in ¹³¹I. Iodine-124¹²⁴I 4.18 d Used as a tracer. Iodine-125 ¹²⁵I 59.4 d Used in cancerbrachytherapy (prostate and brain), also diagnostically to evaluate thefiltration rate of kidneys and to diagnose deep vein thrombosis in theleg. It is also widely used in radioimmuno-assays to show the presenceof hormones in tiny quantities. Iodine-131 ¹³¹I 8.02 d Widely used intreating thyroid cancer and in imaging the thyroid; also in diagnosis ofabnormal liver function, renal (kidney) blood flow and urinary tractobstruction. A strong gamma emitter, but used for beta therapy.Iridium-192 ¹⁹²Ir 74 d Supplied in wire form for use as an internalradiotherapy source for cancer treatment (used then removed). Betaemitter. Iron-59 ⁵⁹Fe 46 d Used in studies of iron metabolism in thespleen. Krypton-81m ^(81m)Kr 13.1 sec ^(81m)Kr gas can yield functionalimages of pulmonary ventilation, e.g. in asthmatic patients, and for theearly diagnosis of lung diseases and function. Lead-212 ²¹²Pb 10.6 hUsed in TAT for cancers or alpha radioimmunotherapy, with decay products²¹²Bi and ²¹²Po delivering the alpha particles. Lutetium-177 ¹⁷⁷Lu 6.7 dIncreasingly important as it emits just enough gamma for imaging whilethe beta radiation does the therapy on small (e.g. endocrine) tumors.Its half-life is long enough to allow sophisticated preparation for use.Molybdenum-99 ⁹⁹Mo 66 h Used as the ‘parent’ in a generator (e.g., a⁹⁹Mo cow) to produce ^(99m)Tc. Nitrogen-13 ¹³N 9.97 m Positron emitterused in PET for studying brain physiology and pathology, in particularfor localizing epileptic focus, and in dementia, psychiatry andneuropharmacology studies. Also has a role in cardiology. Neptunium-238²³⁸N 2.11 d Can be obtained from neutron bombardment of ²³⁷N, a parentof ²³⁸Pu. Oxygen-15 ¹⁵O 122.2 s Positron emitter used in PET forstudying brain physiology and pathology, in particular for localizingepileptic focus, and in dementia, psychiatry and neuropharmacologystudies. Also has a role in cardiology. Palladium-103 ¹⁰³Pd 17 d Used tomake brachytherapy permanent implant seeds for early stage prostatecancer. Phosphorus-32 ³²P 14 d Used in the treatment of polycythemiavera (excess red blood cells) and laboratory experiments. Beta emitter.Plutonium-238 ²³⁸Pu 87.7 y Used as a source in radioisotopethermoelectric generators. Potassium-42 ⁴²K 12 h Used for thedetermination of exchangeable potassium in coronary blood flow.Radium-227 ²²⁷Ra 42 m Parent of ²²³Ra. Radium-223 ²²³Ra 11.4 d Used fortreating pain associated with multifocal bone metastases. Decay productof ²²⁷Ra via ²²⁷Ac and ²²⁷Th. Rhenium-186 ¹⁸⁶Re 3.8 d Used for painrelief in bone cancer. Beta emitter with weak gamma for imaging.Rhenium-188 ¹⁸⁸Re 17 h Used to beta irradiate coronary arteries from anangioplasty balloon. Rubidium-81 ⁸¹Rb 4.6 h Parent of ^(81m)Kr.Rubidium-82 ⁸²Rb 1.26 min Convenient PET agent in myocardial perfusionimaging. Ruthenium - ^(xx)Ru Used in brachytherapy. various isotopesSamarium-153 ¹⁵³Sm 47 h Very effective in relieving the pain ofsecondary cancers lodged in the bone, sold as Quadramet ™. Also veryeffective for prostate and breast cancer. Beta emitter. Selenium-75 ⁷⁵Se120 d Used in the form of seleno-methionine to study the production ofdigestive enzymes. Sodium-24 ²³Na 15 h For studies of electrolyteswithin the body. Strontium-82 ⁸²Sr 25 d Used as the ‘parent’ in agenerator to produce ⁸²Rb. Strontium-89 ⁸⁹Sr 50 d Very effective inreducing the pain of prostate and bone cancer. Beta emitter.Technetium-99m ^(99m)Tc 6.0 h Used in to image the skeleton and heartmuscle in particular, but also for brain, thyroid, lungs (perfusion andventilation), liver, spleen, kidney (structure and filtration rate),gall bladder, bone marrow, salivary and lacrimal glands, heart bloodpool, infection and numerous specialized medical studies. Thallium-201²⁰¹Tl 73 h Used for diagnosis of coronary artery disease other heartconditions such as heart muscle death and for location of low-gradelymphomas. It is the most commonly used substitute for ⁹⁹Tc incardiac-stress tests. Thorium-227 ²²⁷Th 18.7 d Decay product of ²²⁷Acand parent of ²²³Ra. Xenon-133 ¹³³Xe 5.24 d Used for pulmonary (lung)ventilation studies. Ytterbium-169 ¹⁶⁹Yb 32 d Used for cerebrospinalfluid studies in the brain. Ytterbium-169 ¹⁶⁹Yb 32 d Used forcerebrospinal fluid studies in the brain. Yttrium-90 ⁹⁰Y 64 h Used forcancer brachytherapy and as silicate colloid for relieving the pain ofarthritis in larger synovial joints. Pure beta emitter and of growingsignificance in therapy, especially liver cancer.

Current techniques for the production of radioisotopes involve theirradiation of a precursor material in which some of the precursor isconverted into one or more radioisotopes distributed throughout thematerial. This is followed by dissolution of the material and subsequentseparation of the unreacted precursor material from the producedradioisotopes. In currently known techniques for producing radioactiveisotopes, target materials are often sealed in capsules and placed intoirradiation locations. The irradiations can be performed with reactorsor accelerators using a variety of particles and targets. Afterirradiation, the capsules are placed in shielded containers andtransported to chemistry laboratories capable of handling the highactivity of the material for batch dissolution and recovery of theradioactive isotope product or products.

Targetry Coupled Separations

Using currently known techniques, the vast majority of the initialstarting material in irradiations is unreacted and often must bedissolved to allow the irradiation products to be chemically separated.The separation of the minuscule amount of the irradiation product from alarge excess of unreacted starting material often requires multiplepurification routes after dissolution. If additional product is to beformed from nuclear reactions with the remaining starting material, theremaining starting material must be reformed after dissolution into aform suitable for subsequent irradiations. This requires additionalprocessing steps, often with losses in potential product and additionalwaste generation. However, if the dissolved remaining starting materialis not reformed for additional production, that remaining material mustbe disposed of, representing both a loss of potentially usable (andvaluable) material and an additional disposal cost. For isotopicallyrare or enriched materials, this can be a large expense.

The same issue can also arise in the context of nuclear fuelreprocessing. In this context, nuclear fuel, rather than precursorstarting material, can be processed after irradiation in a powergenerating nuclear reactor, nuclear fission test reactor, researchreactor, or teaching reactor, to remove byproducts of the nuclearfission reaction and to reprocess the unreacted nuclear fuel forrecycling and/or reuse. Similar to radioisotope production, currentrecycling of used nuclear fuel from extraction, ion exchange, orelectrochemical methods first requires dissolution of the fuel into asolution. After the fuel has been dissolved, the remaining steps aredone to remove the unwanted byproducts and to reprocess the unreactednuclear fuel back into a suitable form for reuse as fuel.

This disclosure describes systems and methods for creating (throughirradiation) and removing one or more desired radioisotopes from astarting material and further describes systems and methods that allowthe same starting material to undergo multiple irradiations andseparation operations without extensive, if any, damage to its originalform. In one aspect, targetry coupled separation refer to the selectionof a starting material (including selection of the material's physicalstructure) and separation chemistry in order to optimize the recovery ofthe predetermined irradiation product. The disclosure further describeshow with targetry coupled separations, by removing the newly-createdproduct in a way that allows the same starting material to undergo oneor more subsequent irradiations (that is, without having to dissolve orotherwise destroy the material between irradiations), significant costsavings can be achieved using repeated irradiation and separationoperations on the same starting material. The systems and methodsdescribed achieve this with a minimum waste of the starting materialduring the irradiation and product isotope removal and separationoperations. In contrast with the prior art that requires complete orsubstantial dissolution or destruction of the starting material beforerecovery of any irradiation products, the repeated reuse of the startingmaterial allowed by targetry coupled separation represents a significantincrease in efficiency and decrease in cost over the prior art.

One aspect of this disclosure is a system for generating radioisotopesthat includes: one or more containers, including a first container, thefirst container containing source material that includes at least onetarget material; a radiation generator; a radiation bombardment chamberthat receives radiation from the radiation generator in which theradiation bombardment chamber is adapted to hold and expose the one ormore containers to the received radiation, thereby creating at leastsome first radioisotopes that are a direct product of the exposure ofthe target material to the radiation; an insertion component adapted totransfer an extraction material into the first container, therebycontacting the source material within the first container with theextraction material in which the extraction material is selected todissolve, without dissolving the target material, one or more of a firstradioisotope, a second radioisotope that is a daughter product of afirst radioisotope, or both a first radioisotope and a secondradioisotope; and an extraction component adapted to remove extractionmaterial with the dissolved radioisotope from the first containerwithout removing the target material from the first container.

In the system the target material may be a fissionable material and theradiation generator is a neutron generator. The source material may be aporous form with a pore wall width based on a recoil distance of adirect radioisotope product of the fissionable target material. Thesystem may also include a recovery component adapted to receive theextraction material with the dissolved radioisotope from the firstcontainer and recover some of the at least one species of radioisotopefrom the extraction material. The system may also include a conveyancesystem adapted to physically move a container from the radiationbombardment chamber to a second location for interaction with one orboth of the insertion component or the extraction component. Theconveyance system may be further adapted to repeatedly move a containerfrom the radiation bombardment chamber to the second location and fromthe second location to the radiation bombardment chamber. The conveyancesystem may also be adapted to physically move a plurality of containersfrom the radiation bombardment chamber to the second location. Thefissionable target material may include grains containing uranium oxideor uranium metal having an average particle size of less than an averagerecoil distance of ⁹⁹Mo as a fission product of uranium.

In the system, the extraction material may be selected from asupercritical fluid and an aqueous fluid that preferentially dissolvesthe one or more of a first radioisotope, a second radioisotope that is adaughter product of a first radioisotope, or both a first radioisotopeand a second radioisotope. The extraction material may be supercriticalcarbon dioxide containing a ligand that dissolves the one or more of afirst radioisotope, a second radioisotope that is a daughter product ofa first radioisotope, or both a first radioisotope and a secondradioisotope.

The system may automatically perform a radioisotope generation cycle inwhich the system exposes the first container to radiation, transfers theextraction material into the first container, and removes extractionmaterial with the dissolved radioisotope from the first container. Thesystem may also automatically repeat the radioisotope generation cycleon the first container. The system may process a plurality ofcontainers, including the first container, such that each of theplurality of containers is exposed to radiation.

The target material may include one or more of uranium oxide or uraniummetal in the form of powder, salt, cloth, foam or a colloidal suspensionin liquid. The source material may include radium or radiumelectroplated on beryllium. The at least one species of radioisotopegenerated by the system include one or more of ²²⁷Ac, ²¹³Bi, ¹³¹Cs,¹³³Cs, ¹¹C, ⁵¹Cr, ⁵⁷Co, ⁶⁰Co, ⁶⁴Cu, ⁶⁷Cu, ¹⁶⁵Dy, ¹⁶⁹Er, ¹⁸F, ⁶⁷Ga, ⁶⁸Ga,⁶⁸Ge, ¹⁹⁸Au, ¹⁶⁶Ho, ¹¹¹In, ¹²³I, ¹²⁴I, ¹²⁵I, ¹³¹I, ¹⁹²Ir, ⁵⁹Fe,^(81m)Kr, ²¹²Pb, ¹⁷⁷Lu, ⁹⁹Mo, ¹³N, ¹⁵O, ¹⁰³Pd, ³²P, ²³⁸Pu, ⁴²K, ²²⁷Ra,²²³Ra, ¹⁸⁶Re, ¹⁸⁸Re, ⁸¹Rb, ⁸²Rb, ¹⁰¹Ru, ¹⁰³Ru, ¹⁵³Sm, ⁷⁵Se, ²⁴Na, ⁸²Sr,⁸⁹Sr, ^(99m)Tc, and ²⁰¹Tl. The radiation generator used by the systemmay be selected from one or more of Pu—Be sources, ²⁵²Cf sources, sealedtube radiation generators, dense plasma focus devices, pinch devices,inertial electrostatic confinement devices, sub-critical source drivenassemblies, fission reactors, and accelerator spallation devices.

Another aspect of this disclosure is a method for generating ⁹⁹Mo. Themethod includes: providing a source containing a first mass of uraniumin which the source is in a form in which a majority of uranium atomsare within a selected distance from an available surface of the source;exposing the source to neutrons, thereby reducing the first mass ofuranium in the source to a second mass of uranium less than the firstmass and creating at least some atoms of the ⁹⁹Mo radioisotope andthereby also causing at least some of the newly created atoms of the⁹⁹Mo radioisotope to move toward an available surface of the source; andafter exposing the source to neutrons, removing at least some of theatoms of the ⁹⁹Mo radioisotope from the source without substantiallyremoving uranium from the second mass of uranium in the source.

In the method, the removing operation may remove less than 0.1% of theuranium or even less than 0.01% of the uranium from the second mass ofuranium in the source. The providing operation may include providing asource made at least partially from particles containing uranium oxideor uranium metal in which the particles have a particle size based on arecoil distance of ⁹⁹Mo in the source. The method may include enclosingthe source in a neutronically-translucent container. The method mayfurther include exposing the container enclosing the source to neutronsand removing at least some of the atoms of the ⁹⁹Mo radioisotope fromthe container. The method may include selecting an extraction materialthat dissolves atoms of the ⁹⁹Mo radioisotope without changing the phaseof the uranium in the source or selecting an extraction material inwhich atoms of the ⁹⁹Mo radioisotope are more soluble than atoms ofuranium.

The method may include determining the form of the source based on theselected extraction material. The method may also include selecting acombination of a source and an extraction material, wherein thecombination allows ⁹⁹Mo radioisotope to be removed from the source afterexposure to neutrons without substantially affecting the source. Theremoving operation of the method may include passing an extractionmaterial selected to dissolve the ⁹⁹Mo radioisotope through thecontainer, thereby contacting the available surface of the source withthe extraction material.

In the method, the extraction material may be selected from asupercritical fluid and an aqueous fluid. If the extraction material issupercritical carbon dioxide, it may contain a ligand that dissolves the⁹⁹Mo radioisotope. The ligand may be selected from 8-hydroxyquinoline,α-benzoinoxime, disodium 4,5-dihydroxy-1,3-benzenedisulfonate, phosphatecompounds, and diketone compounds. In an alternative embodiment, theligand may have one or more functional groups selected from hydroxyl,carbonyl, diketones, aldehyde, haloformyl, carbonate ester, carboxylate,ester, ether, peroxy, amine, carboxamide, imide, imine, nitrate,cyanate, thiol, sulfide, sulfinyl, sulfonyl, thiocyanate,isothiocyanate, phosphate, and phosphono groups.

The method may include repeatedly performing the exposing operation andthe removing operations on the container without removing the uraniumfrom the container. The method may include removing the ⁹⁹Moradioisotope from the extraction material and, afterwards, repeating theexposing operation on the same source. The method may further includeremoving, in addition to the ⁹⁹Mo radioisotope, an amount of one or moreother fission products created during the exposing operation.

Yet another aspect of this disclosure is a method for selectivelymanufacturing a radioisotope. The method includes: selecting theradioisotope; identifying a target material from which the selectedradioisotope can be created as a fission product; determining a recoildistance of the selected radioisotope in the target material; creating aplurality of grains of target material having a grain size based on therecoil distance of the selected radioisotope; exposing the grains oftarget material to neutrons, thereby causing at least some atoms of thetarget material to undergo nuclear fission to create atoms of theselected radioisotope and also causing at least some of the newlycreated atoms of the selected radioisotope to move the recoil distancerelative to the target material; and extracting atoms of the selectedradioisotope from the target material.

In the method, the exposing and extracting operations may be performedwithout changing the phase of the target material. The method mayfurther include repeating the exposing operation and extractingoperation on the plurality of grains of target material. The pluralityof grains of target material may be contained within aneutronically-translucent container and the method may further includerepeatedly performing the exposing operation and extracting operation onthe same plurality of grains of target material without removing theplurality of grains of target material from the container. The grainsmay be particles of uranium oxide or uranium metal having a grain sizeof less than 20 microns. In an embodiment, the grains are particles ofuranium oxide or uranium metal having a grain size from about 0.001 to10 micrometers. In the method, at least some of the plurality of grainsmay have a characteristic length along at least one dimension smallerthan or equal to the recoil distance.

The extraction operation of the method may include exposing at leastsome of the grains of target material to a solvent that preferentiallyextracts the selected radioisotope from the target material. The methodmay further include processing the grains of target material into asolid, porous source prior to the exposing operation. The processing mayinclude processing the grains of target material into an open-cell foam,an open lattice, an open framework, a ceramic, a cloth, a thin film, amonolayer, a sponge, a nanocage, or a nanocrystal. The processing mayinclude processing the grains of target material into a solid, poroussource having a surface area greater than 10 m²/g as measured byBrunauer, Emmett and Teller (BET) analysis. The processing may includeone or more of sintering, milling, sieving, 3D printing, crystallizing,precipitating, or heating the grains of target material.

Another aspect of this disclosure is a method for selectivelymanufacturing a radioisotope. The method includes: receiving a sourcehaving solid fissionable material in a neutronically-translucentcontainer in which the source has a porous form with pore walls having awidth substantially similar to a recoil distance of a radioisotopeproduct of the solid fissionable material; exposing the source toneutrons, thereby converting at least some atoms of the solidfissionable material via nuclear fission into atoms of the radioisotopeso that the source contains radioisotope and unconverted solidfissionable material; selecting an extraction material thatpreferentially dissolves the radioisotope relative to the fissionablematerial; injecting the extraction material into the container, therebycontacting the source material with the extraction material; removingextraction material from the container after a residence time, therebyremoving at least some dissolved radioisotope from the container whileleaving substantially all of the unconverted solid fissionable materialin the container; and, after removing the extraction material,re-exposing the source material to neutrons, thereby converting at leastsome atoms of the unconverted solid fissionable material via nuclearfission into atoms of the radioisotope.

The method may further include repeating the injecting and removingoperations on the target after re-exposing the target to neutrons. Themethod may also include separating dissolved radioisotope from theextraction material; and incorporating the dissolved radioisotope into adaughter isotope generator. The method may further include periodicallymilking the daughter isotope generator for the daughter isotope. Themethod may include waiting until at least a first predetermined periodof time has elapsed after removing the extraction material from thecontainer in which the predetermined period of time being based on ahalf-life of the radioisotope; and separating the radioisotope from theextraction material.

Another aspect of this disclosure is a method for manufacturing aradioisotope-generating target. The method includes: providing adissolved salt of a fissionable material in a solution in which thefissionable material capable of generating a first designatedradioisotope when exposed to neutrons and the first designatedradioisotope having a recoil distance associated with the fissionablematerial; precipitating an oxide of the fissionable material from thesolution; and selectively forming the precipitated oxide into grains,i.e., individual particles, having a grain size based on the recoildistance of the first designated radioisotope. The method may furtherinclude mixing a precipitant into the solution and/or selecting thefissionable material based on the first designated radioisotope. Themethod may include determining the recoil distance of the firstdesignated radioisotope based on the selected fissionable materialand/or forming grains having a grain size equal to or less than 10micrometers. The method may include forming grains having a grain sizeequal to or less than 1 micrometer. The method may include forminggrains having a grain size equal to or less than 100 nanometers and themethod may include forming grains having a grain size equal to or lessthan 10 nanometers.

In the method, the forming operation may include one or more of milling,drying, filtering, washing, calcining, or sintering the precipitatedoxide. The method may include packaging the grains of the precipitatedoxide in a container. The container may have a first valve adapted toallow the injection of a solvent into the container and a second valveadapted to allow the extraction of a solvent from the container. Thecontainer may be neutronically-translucent. Packaging the grains mayfurther include placing the grains in a cavity defined by the container;and sealing the container, thereby trapping the grains in the cavity.The method may include synthesizing a ceramic from the precipitatedoxide grains. In the method, the providing operation may further includeproviding the dissolved salt of the fissionable material in a solutionselected from one or more of an acidic solution, a basic solution, anaqueous solution, and an alcohol solution.

In yet another aspect of this disclosure, a radioisotope-generatingtarget is described. The target includes a target material capable ofgenerating the radioisotope upon prolonged exposure to neutrons and theradioisotope associated with a recoil distance; and the target materialhaving a characteristic distance selected based on the recoil distanceof the radioisotope. In the target, the radioisotope may be a directfission product of nuclear fission of the target material. Aneutronically-translucent container may be used to contain the pluralityof grains. The container may have an input valve and an output valveallowing the injection and extraction of a fluid. The target's containermay include a body portion and at least one removable lid portion that,when engaged, encloses the target material within the container. Thecontainer may be made of one or more of aluminum, aluminum alloy,zirconium, zirconium alloy, molybdenum, molybdenum alloy, and stainlesssteel.

The target material may include a plurality of grains of target materialloosely packed in the container and the characteristic distance is agrain size selected based on the recoil distance of the radioisotope.The target material may include a plurality of grains of target materialformed into a ceramic. The target material may include a plurality ofgrains of target material formed into or attached to a metal-organicframework. The target material may include one or more of uranium oxideor uranium metal. The target material may include a plurality of grainsof target material formed into a loose powder, a cloth, a foam or acolloidal suspension in liquid. The target material may include radiumor radium electroplated on beryllium. The target material may includegrains of an actinide monolayer and the actinide monolayer may be amonolayer of uranium. The target material may include grains of a highsurface area, uranium metal which may be created using the Krollprocess.

Yet another aspect of this disclosure is a supercritical carbon dioxideseparation method. The method extracts a first radioisotope fromirradiated fissionable material containing a plurality of radioisotopesincluding the first radioisotope. The method includes: selecting aligand that is soluble in supercritical carbon dioxide (sCO₂), forms achelate with the first radioisotope, and does not form a chelate withthe fissionable material; dissolving the identified ligand into sCO₂ toform an sCO₂-ligand solution; contacting the irradiated material withthe sCO₂-ligand solution for a contact time, thereby creating ansCO₂-radioisotope complex solution; separating the sCO₂-radioisotopecomplex solution from the irradiated material; and after separating thesCO₂-radioisotope complex solution from the irradiated material,removing the radioisotope from the sCO₂-radioisotope complex solution.In the method, removing the radioisotope from the sCO₂-radioisotopecomplex solution may include removing the radioisotope complex from thesCO₂-radioisotope complex solution. The removing operation may generatethe sCO₂-ligand solution suitable for reuse without decompressing andrepressurizing the sCO₂ ligand solution. This may be achieved bycontacting the sCO₂-radioisotope complex solution with an acidicsolution, thereby generating an acid-radioisotope solution and aregenerated sCO₂-ligand solution.

In the method, the irradiated material may be enclosed in a containerand the exposing operation may further include passing the sCO₂-ligandsolution through the container without removing substantially any of thefissionable material from the container. In this case, the container maybe operated as a packed bed reactor. In the method, the irradiatedmaterial may be in the form of loose grains and the exposing operationfurther includes passing the sCO₂-ligand solution through the containerat a flow rate sufficient to fluidize the plurality of grains within thecontainer. In the method, the irradiated material may also be a liquid.

In the method, the radioisotope may be ⁹⁹Mo, the fissionable materialmay be ²³⁵U and the ligand may have one or more functional groupsselected from hydroxyl, carbonyl, diketones, aldehyde, haloformyl,carbonate ester, carboxylate, ester, ether, peroxy, amine, carboxamide,imide, imine, nitrate, cyanate, thiol, sulfide, sulfinyl, sulfonyl,thiocyanate, isothiocyanate, phosphate, and phosphono groups. The ligandmay be selected from a fluorinated β-diketone and a trialkyl phosphate,or a fluorinated β-diketone and a trialkylphosphine oxide. The ligandmay be selected from dithiocarbamates, thiocarbazones, β-diketones andcrown ethers. The ligand may have one or more functional groups selectedfrom hydroxyl, carbonyl, diketones, aldehyde, haloformyl, carbonateester, carboxylate, ester, ether, peroxy, amine, carboxamide, imide,imine, nitrate, cyanate, thiol, sulfide, sulfinyl, sulfonyl,thiocyanate, isothiocyanate, phosphate, and phosphono groups.Radioisotopes that may be created by this method include one or more of²²⁷Ac, ²¹³Bi, ¹³¹Cs, ¹³³Cs, ¹¹C, ⁵¹Cr, ⁵⁷Co, ⁶⁰Co, ⁶⁴Cu, ⁶⁷Cu, ¹⁶⁵Dy,¹⁶⁹Er, ¹⁸F, ⁶⁷Ga, ⁶⁸Ga, ⁶⁸Ge, ¹⁹⁸Au, ¹⁶⁶Ho, ¹¹¹In, ¹²³I, ¹²⁴I, ¹²⁵I,¹³¹I, ¹⁹²Ir, ⁵⁹Fe, ^(81m)Kr, ²¹²Pb, ¹⁷⁷Lu, ⁹⁹Mo, ¹³N, ¹⁵O, ¹⁰³Pd, ³²P,²³⁸Pu, ⁴²K, ²²⁷Ra, ²²³Ra, ¹⁸⁶Re, ¹⁸⁸Re, ⁸¹Rb, ⁸²Rb, ¹⁰¹Ru, ¹⁰³Ru, ¹⁵³Sm,⁷⁵Se, ²⁴Na, ⁸²Sr, ⁸⁹Sr, ^(99m)Tc, and ²⁰¹Tl.

Another aspect of this disclosure is a method of obtaining aradioisotope from a bulk material, in which the bulk material includesat least the radioisotope and a fissionable material. The methodincludes: selecting an extraction material that removes the radioisotopefrom the bulk material without substantially dissolving the fissionablematerial; contacting the bulk material with the extraction material fora residence time, thereby creating an extraction material andradioisotope mixture; after the residence time, removing the extractionmaterial and radioisotope mixture; and separating the radioisotope fromthe extraction material. In the method, the contacting operation mayfurther include one or more of: agitating one or both of the bulkmaterial and the extraction material during at least a portion of theresidence time; changing a temperature of one or both of the bulkmaterial and the extraction material during at least a portion of theresidence time; and changing a pressure of one or both of the bulkmaterial and the extraction material during at least a portion of theresidence time.

In embodiments of the method in which the bulk material is solid,contacting the bulk material may include contacting the bulk materialwith a liquid extraction material for a residence time, thereby creatingan extraction material and radioisotope liquid mixture. In embodimentsof the method in which the bulk material is a liquid, contacting thebulk material may include contacting the bulk material with a liquidextraction material for a residence time, thereby creating an extractionmaterial and radioisotope liquid mixture immiscible in the bulkmaterial. In an embodiment, the bulk material may be in the form ofsolid grains stored in a container and the contacting operation mayinclude inserting an amount of the extraction material into thecontainer; and retaining the extraction material in the container forthe residence time.

In the method, the extraction material may include an extractant and asolvent. The extractant may be a ligand soluble in the solvent undertemperature and pressure conditions of the contacting operation. Thesolvent may be sCO2. The ligand may form a carbon dioxide solublechelate with the radioisotope. The ligand may be selected from afluorinated β-diketone and a trialkyl phosphate, or a fluorinatedβ-diketone and a trialkylphosphine oxide or selected fromdithiocarbamates, thiocarbazones, 13 -diketones and crown ethers. Theligand may have one or more functional groups selected from hydroxyl,carbonyl, diketones, aldehyde, haloformyl, carbonate ester, carboxylate,ester, ether, peroxy, amine, carboxamide, imide, imine, nitrate,cyanate, thiol, sulfide, sulfinyl, sulfonyl, thiocyanate,isothiocyanate, phosphate, and phosphono groups.

These and various other features as well as advantages whichcharacterize the systems and methods described herein will be apparentfrom a reading of the following detailed description and a review of theassociated drawings. Additional features are set forth in thedescription which follows, and in part will be apparent from thedescription, or may be learned by practice of the technology. Thebenefits and features of the technology will be realized and attained bythe structure particularly pointed out in the written description andclaims hereof as well as the appended drawings.

It is to be understood that both the foregoing general description andthe following detailed description are explanatory and are intended toprovide further explanation of the invention as claimed.

BRIEF DESCRIPTION OF THE DRAWINGS

The following drawing figures, which form a part of this application,are illustrative of described technology and are not meant to limit thescope of the invention as claimed in any manner, which scope shall bebased on the claims appended hereto.

FIG. 1 illustrates an embodiment of a targetry coupled separation methodthat repeatedly generates radiation products, such as radioisotopes,from the same target.

FIG. 2 illustrates an embodiment of a targetry coupled separation systemfor the continuous or semi-continuous production of the ⁹⁹Moradioisotope.

FIG. 3 illustrates an embodiment of a method for selectively generatinga desired radioisotope using targetry coupled separation.

FIG. 4 illustrates an embodiment of a container suitable for holdingtarget material in a targetry coupled separation.

FIG. 5 illustrates an embodiment of a method of manufacturing aradioisotope-generating target in greater detail.

FIGS. 6A through 6C illustrate a more detailed means for characterizinggrain size of a granular target material.

FIG. 7 illustrates an embodiment of general separation method suitablefor use with targetry coupled separation.

FIG. 8 illustrates an embodiment of a method for the reformation ofnuclear fuel using supercritical carbon dioxide (sCO₂).

FIG. 9 illustrates the chemical structure of the ligand,hexafluoroacetylacetonate (“hfac”), suitable for use in targetry coupledseparation with some specific radionuclides.

FIG. 10 illustrates an embodiment of a method of extracting a firstradioisotope product from irradiated fissionable source material.

FIG. 11 illustrates the material conversion cycle showing the changes ina source over two passes of targetry coupled separation.

FIG. 12 illustrates an alternative embodiment of a method forselectively generating a desired radioisotope using targetry coupledseparation.

DETAILED DESCRIPTION

FIG. 1 illustrates, at a high level, an embodiment of a targetry coupledseparation method that repeatedly generates irradiation products, suchas radioisotopes, from some amount of target material. In the method 10as illustrated, some amount of target material is irradiated in anirradiation operation 12 which creates a desired irradiation product.The irradiation operation 12 is followed by a separation operation 14 inwhich the desired product is removed from the target material withoutsubstantially reducing the amount of post-irradiation target material.(As discussed in greater detail below, the word ‘substantially’ shall beused at times when referring to the amount of target material thatremains in the irradiated object after a separation operation 14 toremind the reader that no separation technique is perfect and a small orde minimis amount of the target material may, in fact, be removed duringthe separation operation 14.) The desired product, after removal fromthe target material, may then be subjected to subsequent processing anduse. For example, in an embodiment the target material is incorporatedinto a porous, solid object and the product is removed using a liquidsolvent that dissolves the product but does not substantially, if atall, dissolve or remove the remaining post-irradiation target materialfrom the object. The irradiation and separation operations 12, 14 arethen repeated on the remaining target material. The method 10 may berepeated any number of times and, in an ideal system, could be repeateduntil all of the target material is completely consumed. Realistically,however, it is presumed that after some number of repetitions it willbecome more economical to dispose of the remaining target materialrather than reuse it for another cycle. While many more detailedembodiments, some of which are discussed below, are possible, FIG. 1 ispresented as a simplified embodiment in order to provide a convenientreference point for further discussion and to introduce the concepts andterminology that will be discussed in greater detail below.

As discussed above, in targetry coupled separations a target material issubjected to one or more irradiation operations. The “target material,”as that term will be used herein, refers to a material that, uponexposure to the particular radiation used in an irradiation operation,results in the creation of one or more irradiation products. Dependingon the embodiment, the radiation used may include one or more of alphaparticles, beta particles, gamma rays, x-rays, neutrons, electrons,protons, and other particles capable of producing nuclear reactionproducts. In any particular irradiation operation, some amount of thetarget material will be converted into the irradiation product(s),resulting in a mass decrease of target material and a newly created massof the irradiation product(s).

In some embodiments, targetry coupled separations may be tailored toenhance the recovery of one or more predetermined, desired products fromthe irradiated target material. A desired product refers to either adirect or an indirect irradiation product that the operator wants toremove from the target material after irradiation in the separationoperation 14. Depending on the combination of radiation and targetmaterial used in an embodiment, undesired reaction products may also becreated by the irradiation, which may not be removed from the targetmaterial in the separation operation 14. For example, if the radiationis in the form of neutrons and the target material includes uranium-235(²³⁵U), one of the fission products will be the molybdenum isotope,⁹⁹Mo. After irradiation, atoms of ⁹⁹Mo will be dispersed within thetarget material and each ⁹⁹Mo atom will be from one uranium atom thatexisted prior to irradiation. However, due to the nature of neutronirradiation, many other fission products will also exist in the targetafter irradiation, each also representing atoms produced from fissioneduranium atoms. In an embodiment, the ⁹⁹Mo is a desired product andsubsequently removed in the separation operation 14 while the otherfission products are not removed and remain with the target materialduring subsequent irradiations.

Target material may be incorporated into a larger mass of sourcematerial. The source material may be formed into a single object ordiscrete mass, occasionally referred to herein simply as the “source”,that can be exposed to radiation in an irradiation operation to convertat least some of the target material (either directly or indirectly, asdiscussed in greater detail below) into the desired product (or itsparent, as will be discussed below). In addition to the target material,source material also may optionally include material that does not reactto irradiation to produce the desired product. Such material may becompletely unreactive to the radiation or may form something other thanthe desired product. Where appropriate, the term ‘ancillary material’may be used to refer to any component of the source material that doesnot form the desired product when irradiated. Ancillary materials couldinclude, for example, trace contaminants, materials present in thesource material to provide a physical structure for the target, orunharvested products of previous irradiations of the target.

FIG. 11 illustrates the material conversion cycle showing the changes ina source over two passes of targetry coupled separation. The materialconversion cycle 1100 starts with some amount of source material 1102that, in the embodiment shown, includes some amount of target 1120 andanother amount of ancillary material 1122. Although the target andancillary material of the source material are shown as separate boxes,it is to be appreciated that the FIG. 11 is for illustrative purposes ofthe cycle of target and product, and it should be appreciated that theelements of FIG. 11 are not representative of actual amount or ratio oftarget and ancillary material, intermixing of target and ancillarymaterial, and/or structure of the source. An irradiation operation (suchas irradiation operation 12 of FIG. 1) changes the source material 1102into an irradiated source material 1104 in which some amount of targetmaterial 1120 has been changed into irradiation products 1124. Whileexaggerated for illustration purposes, in FIG. 11 approximately half ofthe target material 1120 has been changed into some amount ofirradiation product 1124. Because products 1124 are considered anancillary material, FIG. 11 also illustrates a relative increase in themass of ancillary material in the source 1104 and a commensurate deceasein the amount of target material 1120.

The reader will be reminded that, especially in fission reactions, theimmediate result of irradiation will be a spectrum of direct irradiationproducts, some which over time may subsequently decay into indirectproducts which may, themselves, further decay into other indirectproducts. Thus, the exact makeup of irradiation products 1124 may changeover time as various direct and indirect products decay. However, forthe purposes of this discussion, FIG. 11 does not distinguish betweendirect irradiation products and indirect irradiation products or attemptto track how the makeup of irradiation products changes over time.

The cycle 1100 further shows the effects of a first separation operation(such as separation operation 14 of FIG. 1) on the irradiated sourcematerial 1104. The separation results in a certain amount of desiredproduct 1126 being removed from the irradiated source material 1104.Again, exaggerated for illustration purposes, FIG. 11 shows thepost-separation source 1106 having had some of the product removed, sothat the post-separation source material 1106 has relatively lessancillary material, but the amount of target remains the same as in theirradiated source material 1104. This graphically illustrates that theseparation operation has no effect or substantially no effect on themass of target material 1120 in the source.

FIG. 11 also illustrates that some irradiation products 1124 may remainas ancillary material 1122 in the source 1106 after the separation. Thismay be the case either because the separation is not 100% efficient,because not all of the irradiation products 1124 are desired productsand the separation operation intentionally does not remove thoseproducts, or both.

FIG. 11 further illustrates a second set of irradiation and separationoperations on the source material. FIG. 11 shows a second-irradiatedsource 1108 that again illustrates that some amount of target material1122 of the precursor source material 1106 is converted into product1124 by the second irradiation operation. The second separationoperation then reduces the overall mass of the second-irradiated sourcematerial 1108 by removing some of the desired product 1126, but withoutchanging the mass of the target material 1122 in the source material1108. The resulting post-second separation source material 1110 is thenready for subsequent irradiation and separation operations asillustrated by the arrow at the bottom of the illustration.

As mentioned above, FIG. 11 is exaggerated for illustration purposes.However, it clearly shows certain aspects of targetry coupledseparation. Specifically, it illustrates that target material 1122 isconverted into product 1124 by the irradiation operation and some amountof product 1126 is removed, without removing substantially any targetmaterial 1122 from the source, in the separation operation. Thus, bysubjecting the same source to repeated irradiation and separationoperations, the target material 1122 in the source can be consumed untilsuch time as it is completely converted into product 1124 or it is nolonger economical to repeat the process.

FIG. 11 further illustrates that not all of the product 1124 may beremoved by the separation operation. This may occur for differentreasons. While it is preferable to remove as much of the desired product1126 as possible with each separation, not all of the irradiationproducts 1124 may be desired products 1126 and/or removal of all of thedesired product 1126 may not be technologically practical or possible.Thus, products 1124 from prior irradiation operations (such as undesiredproducts) may remain in the source material by design (e.g., byappropriate selection of the extraction material to avoid or reduceremoval of the undesired product). It is also possible that theseparation operation is not 100% efficient at removing all the product,thus leaving some desired product in the source material.

As discussed above, target material can include any one or more isotopesor elements that, directly or indirectly, can form the desired productupon irradiation. The term ‘directly or indirectly’ is used here topoint out to the reader that, while some desired isotopes may be thedirect irradiation product of a target material, other desired productsmay be created by the natural decay of a direct irradiation product. Forexample, ⁹⁹Mo is one of many direct fission products of ²³⁵U. That is,in the thermal neutron fission of a mass of ²³⁵U, some of the atoms(6.1% to be precise) of ²³⁵U will be converted directly into atoms witha mass of 99, including ⁹⁹Mo. Other atoms of ²³⁵U will be converted intoother products such as ¹³⁵I and ¹⁵⁷Gd. However, many direct fissionproducts are unstable and will, after some period of time based on theirhalf-lives, naturally decay into indirect products. Using ⁹⁹Mo again asthe example, ⁹⁹Mo has a half-life of 65.94 hours, primarily decayinginto ^(99m)Tc. The isotope ^(99m)Tc, with a 6.01 hour half-life, decaysinto ⁹⁹Tc. Thus, ⁹⁹Mo is a direct product of the fission of ²³⁵U while⁹⁹Tc is an indirect product. It should be noted that ⁹⁹Tc is also adirect product of fission, but with different independent fission yieldthan ⁹⁹Mo. It should be noted that a desired product may be both adirect product from irradiation of a target and an indirect product thatis created by decay of a different direct product of the sameirradiation of the target. Target material can include the elementalform of a material, metals, alloys, intermetallic compounds, hydrides,oxides, hydroxide, halides, chalcogenide, nitrides, phosphides,carbides, silicides, carbonates, nitrates, sulfates, thiosulfate,sulfites, perchlorates, borides, arsenates, arsenites, phosphates,nitrite, iodate, chlorate, bromate, chlorite, chromate, cyanides,thiocyanates, amides, peroxides, organic complexes, mixed species,ternary compounds, quaternary compounds or greater, or a combination ofany of these compounds.

The source material can be in a variety of structures, forms ormorphologies that permit the separation of the desired product from thetarget without significant alteration of the physical form of the sourcematerial (other than the removal of some or all of the desiredproducts), thus allowing previously irradiated source material toundergo a subsequent irradiation without substantial reprocessing.Morphologies, forms, and shapes can include sheets, monoliths, sol-gels,ceramics, polymers, metallic phases, particles, spheres, layers,aggregates, crystalline phases, metal-organic frameworks, fibers,precipitates, tubes, micelles, sponges, cages, powders, granules,suspensions, slurries, emulsions, porous particles, and colloids.

Furthermore, as will be described in greater detail below with referenceto FIGS. 4 and 5, a source material's physical form or morphology may beselected or altered in order to improve the performance or efficiency ofthe separation operation 14, e.g., by tailoring the form of the targetmaterial in the source to suit the selected extraction material orprocess. For example, a particularly high surface area form of sourcematerial may be used to improve the contact between a solid targetmaterial and a liquid or gaseous extraction material, such as asupercritical carbon dioxide and ligand mixture. Alternatively, a formof source material may be selected to take advantage of the effect ofirradiation. For instance, some uranium fuels (e.g., ceramic and metalfuels) can become porous after irradiation in a reactor, which canprepare the target for separation of the product from the target andsubsequent re-irradiation without the need to dissolve or destroy mostor all, if any, of the remaining source material as part of theseparation.

Although, in an embodiment, the source may be a solid piece or structureincluding the target material, in many embodiments discussed herein thesource material may be contained in a container that, at leastpartially, encases the source material. For example, in an embodimentthe source material may be in a particulate or pelletized form and acontainer may be provided to hold the source material during some or allof the operations of FIG. 1. Depending on the physical form of thesource material (e.g., aggregate, powder, liquid, etc.), a container maybe used to provide a physical constraint and may also be used to providecontact points for ease of handling. In addition, the container may beadapted to simplify the separation operation 14. Suitable containerembodiments are discussed in greater detail with reference to FIG. 4.The form of the porous source material can be manufactured or selected,such as through 3D printing, foam, molds, particulate, sinteringparticulate, etc. as will be discussed further below.

Returning to FIG. 1, in an embodiment of the irradiation operation 12,one or more sources are exposed to radiation that causes at least someof the target material to be converted into desired product. Radiationgenerators can include reactors, particle accelerators, electronaccelerators, plasma focus devices, pinch devices, and/or sealed tubeneutron generators. The accelerators can supply reaction particlesdirectly or can be used to produce particles from reactions. In anembodiment, the irradiation operation 12 may include placing one or moresources containing target material in a controlled environment where thesource(s) may be safely exposed to the radiation. For example, in anembodiment in which the radiation includes neutrons, exposure isachieved by placing the source material in, or passing a source through,a neutron bombardment chamber that receives neutrons from a neutrongenerator.

In the separation operation 14, the exposed source material is treatedto remove the desired product without substantially dissolving orremoving the remaining target material in the source. In an embodiment,this may include contacting an available surface of the source materialwith an extraction material, such as a fluid, that preferentiallydissolves the desired product but for which the target material andancillary material, if any, is either insoluble or has a substantiallyreduced solubility relative to the desired product. In an alternativeembodiment, some other separation technique may be used thatpreferentially removes the desired product from the source material. Thetarget material in a source is left in a form suitable for subsequentirradiation to generate additional desired product.

The reader will appreciate that no separation system is perfect and thatsome trace amount of target material may be unintentionally entrained,dissolved and/or otherwise removed with the extraction material duringthe separation operation 14. As mentioned above, the word‘substantially’ shall be used at times when referring to the amount oftarget that remains in the source material after a separation operation14 to remind the reader that some small or de minimis amount (less than0.1% although less than 0.01% is anticipated) of the target material bymass may, in fact, be removed from the source during the separationoperation 14.

Although the techniques introduced above and discussed in detail belowmay be implemented for a variety of desired products such asradioisotopes or other fission products, this disclosure will primarilydiscuss targetry coupled separation systems and methods in the contextof systems and methods that repeatedly generate and remove one or morefission products from source material containing fissionable material asthe target. More particularly, this disclosure will primarily discusstargetry coupled separation in the context of systems and methods thatrepeatedly generate and remove ⁹⁹Mo as the desired product from grainsof a source material that includes ²³⁵U as the target. Upon fission ofthe uranium, ⁹⁹Mo product is one of the many isotopes produced asfission products. The ⁹⁹Mo product can be separated from the uranium bythe formation of a molybdenum-specific species that can be easilyremoved from the uranium target without the need to remove the targetmaterial from the source or alter the form of the target material tofacilitate separation. An example of suitable molybdenum specie thatfacilitate separation from the source includes MoO₄ ²⁻, which can beremoved by dissolution, or Mo(CO)₆, which can be removed byvolatilization.

The reader will understand that the technology described in the contextof ⁹⁹Mo could be adapted for use in generating any nuclear reactionproduct, such as those listed in Table 1, either directly by irradiatingan appropriate target material with neutrons, or indirectly byirradiating the appropriate target material with neutrons to form aradioisotope parent of the desired product and allowing the parent todecay. More generally, the targetry coupled separation methods andsystems described herein may be adapted to generate any desired productthat could be obtained through irradiation of a target material usingany type of radiation, not just neutron irradiation.

FIG. 2 illustrates, again at a high level, an embodiment of a targetrycoupled separation system. The system illustrated is adapted for thecontinuous or semi-continuous production of products such as the ⁹⁹Moradioisotope from a target containing ²³⁵U.

While embodiments of the system 200 may include manual operations, thesystem 200 is particularly suited for automation and the entire processmay be implemented as an automated system that continuously orsemi-continuously generates ⁹⁹Mo product until such time as the targetis consumed or otherwise fouled with unwanted byproducts to the extentthat further generation of ⁹⁹Mo from the targets is uneconomical. Forexample, some fission products of 235U are neutron poisons (such as¹³⁵Xe, ¹⁴⁹Sm and ¹⁵¹Sm) and, if these products are allowed to buildup inthe source material over successive re-irradiations, the subsequentyield of desired product from each irradiation will be reduced. Eventhen, in an embodiment, old source material may be automatically storedand new source material placed into the system until all available or adesired amount of target material is consumed.

The system 200 includes: a neutron generator 202; a neutron bombardmentchamber 204; a conveyance system 206 (illustrated as a conveyor-typesystem 206); a separation system 208, which in this embodiment includestwo components: an insertion component 210 and an extraction component212; an optional treatment system 228; a product storage system 224; anda supply or source of extraction material 226. A plurality ofuranium-containing sources 214 is illustrated undergoing variousoperations by the separation system 200 and traveling in the directionof the conveyor-type system 206 as indicated by arrows 220 and 222.

In the embodiment shown, the neutron generator 202 can be anyappropriate generator of neutrons. Examples include Pu—Be sources, ²⁵²Cfsources, sealed tube neutron generators, dense plasma focus device,pinch devices, inertial electrostatic confinement device, fissionreactors, or accelerator spallation devices.

The neutron bombardment chamber 204 receives neutrons from the neutrongenerator 202 and exposes any sources 214 within the chamber 204 toneutron bombardment. The chamber 204 may include multiple componentsdesigned to allow the sources to enter and exit. The chamber 204 may beconstructed to reduce the release of stray neutrons to the outsideenvironment of chamber 204 itself or outside of the containment ofsystem 200. The chamber 204 may include an irradiation zone within whichthe sources are exposed to neutrons. The irradiation zone may be sizedto irradiate any desired number of sources at the same time. In theembodiment shown, the conveyor 206 causes sources to pass through theirradiation zone.

Because the rate of source transport into and through the irradiationzone determines, in part, the total exposure of the source to neutrons,the transport rate may be selected to achieve the desired amount ofirradiation of the target based on the neutron flux of the neutrongenerator 202. Transportation of sources may be continuous (wherein thesources are continuously in motion), non-continuous (in which theconveyance system 206 starts and stops to achieve the desired rate), ora combination of the two (e.g., continuous movement through thebombardment chamber 204 but sources are held in the separation system208 until a desired amount of separation has been obtained after whichtransportation is resumed). The rate may be constant, decreasing,intermittent or varied based on monitoring of neutron flux or any otherparameter that can be used to identify the exposure of the targetmaterial to radiation. It should be appreciated that the exposure levelof irradiation in the irradiation zone need not be constant for aparticular source, or from one source to the next source introduced tothe irradiation zone. That is, the neutron generator 202 may not have aconstant neutron flux over time. In this situation, the neutron flux maybe monitored and the transport rate may be varied as necessary toachieve the desired irradiation results.

When a new source enters the irradiation zone, there is little or noproduct or an undesired amount of product in the source material. Giventime in the irradiation zone, the nuclear reaction product concentrationincreases in the source as some of the atoms of uranium undergo anuclear reaction due to the uranium atoms' interaction with theneutrons. The conveyor speed, stopping points and times and/or neutronflux in the irradiation zone may be tuned so that sources are exposedfor a desired irradiation time or dosage, thus generating a designedamount of fission products including ⁹⁹Mo in each source. The source isthen removed from the irradiation zone by further movement of theconveyor 206 and passed to the separation system 208.

The separation system 208 refers to those components that together passextraction material through the source to remove at least some of theatoms of the ⁹⁹Mo radioisotope product from the source withoutsubstantially reducing the post-irradiation uranium content of thesource. The separation system 208 obtains extraction material from theextraction material supply 226, contacts the irradiated source materialwith the extraction material, and then removes the extraction material(along with at least of the ⁹⁹Mo product) for the source material. Asillustrated graphically in FIG. 11, the separation system 208 does notsubstantially, if at all, reduce the mass of the target material in thesource, but rather exclusively or primarily removes only the desiredproduct or products. Generally, an extraction material may be introducedto the irradiated source material (including the product material) todissolve the ⁹⁹Mo radioisotope product into the extraction material andsubstantially retain the full, post-irradiation mass of target materialin the source material separate from the extraction material. Thechemistry of the extraction material used to perform the separation maybe tailored to the target material and the desired product, and,depending on the embodiment, may use aqueous solutions, organic phases,ionic liquids, supercritical fluids, fluidized beds, reactive gases,thermal treatments, or their combinations. However, in this embodimentthe separation system 208 uses an extraction material that is placed incontact with the irradiated target material, now containing an amount of⁹⁹Mo product. In an embodiment, the extraction material preferentiallydissolves the ⁹⁹Mo product without dissolving the uranium target or,preferably, any of the ancillary material including any other byproductssuch as other fission products. In an alternative embodiment, thedesired product is, in fact, multiple fission products and theextraction material preferentially dissolves all of the desired productssimultaneously, without substantially affecting the remaining targetmaterial in the source material. In yet another embodiment, multipledifferent extraction materials are used sequentially in separatecontacting operations to remove the various products. In yet anotherembodiment, multiple different extraction materials are used in a singlecontacting operation to remove the various desired products.

The removed product(s) are then recovered from the extraction material(s) and processed as necessary into a usable form and the sources arereturned to the conveyance system 206 for further irradiation. In theembodiment shown, the product is output into a product storage system224. The recovery of the product(s) may be performed by the extractionsystem 208 so that a final, usable form of product(s) is stored by thestorage system 224. In an alternative embodiment, the product andextraction material mixture may be stored in the storage system 224 forfuture processing by a separate recovery system (not shown), which maybe local to or remote from the system 200.

In the embodiment illustrated in FIG. 2, the first stage of theseparation system 208 is the insertion component 210. In the embodimentshown, the targets are contained in containers 214 encasing some amountof target material and the insertion component 210 refers to thatequipment that transfers the extraction material into the containers.The insertion component is adapted to transfer an extraction materialinto containers, thereby contacting the source material within the firstcontainer with the extraction material. In an embodiment, the extractionmaterial is selected to dissolve, without dissolving the targetmaterial, the desired product or products. Transferring the extractionmaterial into the containers may include one or more of injecting theextraction material under pressure into the container, applying a vacuumto a container open to a reservoir of extraction material, allowing theextraction material to flow under gravity into the container, submergingan open container into a pool of extraction material, or any othertechnique whereby the extraction material is transferred into thecontainer. The insertion component 210 may include automated ormanually-operated equipment that accesses the container and delivers theextraction material into the container, such as through one or morevalves or other access points provided on the container. As is known inthe art, there are many different ways of inserting fluids into acontainer and any suitable method may be used.

In an embodiment, the extraction material is maintained in the containerfor an appropriate residence time. During some or all of the residencetime, the container may be subjected to additional actions such asheating, cooling, pressurization, depressurization, agitation,circulation of extraction material, and/or secondary irradiation asdesired to improve the removal of the product from the source material.For example, in an embodiment the source material is a loose particulateor powder and the extraction material is repeatedly flowed (circulated)under pressure through the container (e.g., flowed into a valve at oneend of the container and removed from a valve at the other end of thecontainer) such that the container temporarily becomes a packed bedreactor or, if the flow rate through the container is sufficient, afluidized bed reactor. In these embodiments, the contacting of theextraction material with the source material is performed substantiallywithout removing target material from its container, and in some caseswithout removing any source material other than the desired product fromthe container.

After the appropriate residence time, the extraction component 212 ofthe separation system 208 removes the extraction material from thecontainer and passes the extraction material including the removedproduct to a treatment system 228. As with the insertion component 210,any suitable technique for removing the extraction material and productmixture including those described above for inserting the extractionmaterial into the container may be used. The extraction solution canadmix or carry the product. Alternatively, the extraction solution(including the extraction method and parameters of operation) can beselected to dissolve the product.

The treatment system may separate the dissolved ⁹⁹Mo product from theextraction material. The treatment system 228 and/or a post-processingsystem (not shown) that is considered a part of the separation system208 for the purposes of this discussion may purify the removed productinto a usable ⁹⁹Mo or further decay product , which is then stored inthe product storage system 224. For example, in an embodiment the ⁹⁹Moproduct may be incorporated into an isotope generator by the separationsystem 208 as a final processing step. The extraction material may befurther regenerated for reuse, such as by removal of any unwantedbyproducts or trace source material picked up by the extractionmaterial. However, regeneration is optional and the separation system208 may or may not regenerate the extraction material as part of therecovery of the ⁹⁹Mo product. The extraction material may be returned tothe extraction material supply 226 for reuse by the insertion component210. Alternatively, the extraction material may be processed for wasteand/or removal from the system 200. For example, in an embodiment thatuses sCO₂ as part of the extraction material, the treatment system 228may maintain the sCO₂ in the supercritical state during the separationand returned recycled sCO2 to the separation system 208.

The purification of ⁹⁹Mo product may include removal of a trace amountof target elements or isotopes, removal of other products from thenuclear reaction, and/or removal of separation chemical(s) used in theseparation of the product from the source material. The methods forpurification are based on existing techniques and can include any one ormore appropriate techniques including column chromatography, gravityseparation, distillation, evaporation, centrifugation, precipitation,ion exchange, sorption, filtration, and solvent extraction. Thesemethods can be performed with an automated chemistry system.

Additionally, the extraction component 212 may also perform one or moreregeneration operations to prepare the source material for furtherirradiation. Such regeneration operations can include washing theremaining source material with a volatile, acidic or basic solution,heating, treatment under vacuum, sparging with gas, flushing with asolution, or any other appropriate process or a combination of any ofthese processes. The regeneration operations can occur at the samelocation as the extraction operation and may use the same equipment, asshown in FIG. 2. For example, the extraction component 212 may performthe source regeneration and, in that capacity, may also be considered asource regeneration component. In an alternative embodiment (not shown),the source regeneration operations can occur at a different locationand/or use separate equipment, such as an independent sourceregeneration component (not shown).

Some embodiments of targetry coupled separation may have advantagescompared to the existing methods of isotope production. The reuse of thesource containing the target is an attribute in this regard. Becausetargets may be composed of enriched or rare isotopes, embodiments mayprovide a ready route to re-irradiate the target with reducedpreparation and/or regeneration expenses. Furthermore, in variousembodiments separating the produced isotope product from the source doesnot substantially reduce the amount of target material in the source(after conversion of some amount of target material into product throughirradiation) nor even require the target material be removed from thesource material or even the container. As will be appreciated, targetdissolution can result in waste formation, which can represent asignificant expense with radioactive material. While the target materialmay be recovered after dissolution and reformed into a new source,losses of target through imperfect reformation and/or costs ofreformation may have an impact on fabrication costs and waste formation.And that is not to mention the extra cost associated with reformation ofthe target material into a new source.

Embodiments of targetry coupled separation can be incorporated intoexisting reactors or accelerator centers, thereby utilizing currentinfrastructure which often are included in and/or accompany thesefacilities. This utilization can help to decrease potential productionstart-up and/or change costs for existing irradiation facilities and canhelp to result in a broader distribution of isotope production centers.

In addition, automated or manual embodiments of the system 200 easilymay be installed into existing equipment or installations. For example,embodiments may be incorporated into existing irradiation facilitieswhich may include any component of or combination of equipment toperform and/or support a reactor, accelerator center, target/productchemical processing equipment, etc. Embodiments can be combined withparticle accelerators or reactors to produce desired isotopes.Accelerators and reactors produce different isotopes for a range ofdiagnostic and therapeutic medical applications as well as industrialusage. Through adjustments of the target material, its morphology, andthe separation chemistry, embodiments can be tuned to produce a range ofproduct isotopes for medical applications in the same facility orsimilar facilities to those existing. Embodiments can incorporateexisting chemical automation tools. These automation tools can beapplied to the separation of the produced radionuclide product from thetarget, purification of the separated radionuclide, and any preparationand/or regeneration of the source prior to re-irradiation.

The final radioisotope product of the system 200 can be integrated intoexisting generators. These generators can be distributed to medicalfacilities to provide radionuclides for medical applications. Theprocesses used by the separation system 208 can be selected to regulatethe chemistry and solution conditions of the produced isotope to meetdesired conditions for generator use.

Because radioisotope products are time-sensitive due in part tohalf-life limitations of the produced species, producing product nearits preparation or ultimate use location can increase their availabilityfor medical or other applications. Additionally and/or alternatively,the ability to automate separations and isotope production can increaseproduction rates and help to decrease potential worker dose. The highactivity of targets and/or waste may result in radiation doses toworkers involved in their handling. Thus, embodiments of the system 200and method 10 can couple shorter irradiation time with automation forseparation and/or reformation and reduce waste processing with targetre-use, thereby helping to decrease the potential worker dose due tomaterial handling.

In the embodiment illustrated in FIG. 2, the conveyor 206 is theconveyance system that physically moves the sources from the neutronbombardment chamber to a different location for interaction with one orboth of the insertion component 210 or the extraction component 212.Conveyance system 206 may be open to the environment of other componentsof the system 200 or alternatively enclosed and possibly shielded toreduce radiation emissions around conveyance system 206. For example, asillustrated in FIG. 2 a portion of the conveyance system 206 such as aconveyor belt may physically move through some or all of the othercomponents and systems. Alternatively, the conveyance system 206 maysimply transfer containers 214 between the various components andsystems, each of which is provided with its own container handlingmechanisms for receiving the containers from, and returning them, to theconveyance systems. In an alternative embodiment, other conveyancesystems than conveyors may be used such as robotics, or any othersuitable container handling or transfer system including withoutlimitation belts, chutes, diverter gates, bucket elevators, pneumaticconveyances, screw conveyors, etc.

The conveyor 206 may be operated in a semi-continuous fashion (e.g.,periodically pausing while containers are being acted on by a system orcomponent) or a continuous fashion. In an alternative embodiment, thesystem 200 may produce the product through a batch irradiation followedby a batch separation, for example, irradiation of multiple sourcesand/or containers as a batch. The irradiated containers may be processedfor product extraction either serially or in one or more batches or setsof containers. Although FIG. 2 shows a substantially continuousirradiation followed by a batch separation of the individual containers,any combination of batch or substantially continuous irradiation andbatch or substantially continuous separation may be used as appropriate.

Many different configurations of the targetry coupled separation systemare possible and all are considered within the scope of this disclosure.For example, in an embodiment, the conveyor 206 may eliminated in favorof a manual transfer operation. In this embodiment, operators manuallyor by remote control move the sources between the various components ofthe system 200. In yet another embodiment, the various components of thesystem 200 are designed so that the source is not moved, but rather thedifferent components interact with a stationary source at differenttimes during the process. In yet another embodiment, one or more sourcesare fixed inside a mobile neutron bombardment chamber 204 and thechamber is moved between a neutron generator and a separation system208.

FIG. 3 illustrates an embodiment of a method for selectively generatinga desired radioisotope using targetry coupled separation. The method 300begins with the selection of the radioisotope to be created. This isillustrated by a selection operation 302. In the selection operation 302any radioisotope may be selected, for example from Table 1 above, suchas ⁹⁹Mo, ²³⁸U, ¹³¹I, ⁵¹Cr, ²²⁷Ra, ²²³Ra, ²²⁷Ac, etc. that the operatorultimately wants to obtain. In an embodiment, more than one radioisotopemay be selected.

As already noted, some desirable radioisotopes may not be directproducts of an irradiation operation. In those situations, the selectionoperation 302 may be equally considered a selection of the decay chainor a selection of any of the radioisotopes in the decay chain. Forexample, to obtain ²²³Ra one may wish to create ²²³Ra generator out of²²⁷Ac, as is known in the art. However, for the purpose of thisdisclosure, the term ‘selected radioisotope’ refers to the radioisotopethat is the direct product of the irradiation of the target in theirradiation operation, and the selected radioisotope can be processed(including providing a holding time for anticipated decay) as necessaryto ultimately generate the desired product.

For example, if one wished to use targetry coupled separation toultimately create ²²³Ra for medical use, the selected radioisotope ordirect product would be ²²⁷Ac, e.g., for subsequent incorporation into a²²³Ra generator, the selected radioisotope would be would be ²²⁷Ac.Likewise, if one wished to use targetry coupled separation to generate⁹⁹Mo for subsequent incorporation into a ^(99m)Tc generator, theselected radioisotope would be ⁹⁹Mo (because it is a direct product).However, in this instance, there are also many direct products withatomic number 99 that are parents of ⁹⁹Mo, by one decay chain oranother, and that relatively quickly decay into ⁹⁹Mo. These directproduct parents include: ⁹⁹Nb which decays into ⁹⁹Mo and has a half-lifeof 15 seconds; ⁹⁹Yr which decays into ⁹⁹Nb and has a half-life of 1.47;and ⁹⁹Zr which has a half-life of 2.2 seconds and decays into ⁹⁹Y, tomention just a few in one particular decay chain. Thus, to use targetrycoupled separation to generate ⁹⁹Mo for a ^(99m)Tc generator, in anembodiment the selected radioisotopes may include some or all thosedirect products with atomic number 99 that decay into the desiredproduct of ⁹⁹Mo.

After the radioisotope or isotopes are selected, a target material isidentified from which the selected radioisotope(s) can be createdthrough irradiation. This is referred to as the target identificationoperation 304. The target identification operation may further includeidentifying the overall source material (target and the ancillarymaterial) including the physical properties of the source. In thismanner, the target material identification operation may be referred toas the source material identification operation. For example, if ⁹⁹Mo isa desired product and ⁹⁹Mo and its atomic number 99 direct productparents are the selected radioisotopes, then one suitable targetmaterial may be created from ²³⁵U, such as an oxide of ²³⁵U or pure ²³⁵Umetal, from which ⁹⁹Mo can be obtained directly and indirectly throughneutron bombardment. Many radioisotopes may be obtained from differentcompounds, e.g., from ²³⁵U or ²³⁹Pu and a combination of compounds maybe selected as the target material.

The identified target material may include any fissionable material, orcombination of fissionable materials, or other isotopes suitable forproduction of desired isotopes by nuclear reactions, and may be selectedbased on the type of radiation generator, bombardment chamber, spectrumof the reactor (thermal or fast), and other equipment available. Forexample, the target may incorporate any known material which can befissioned with a neutron to create the direct, selected radioisotopeproduct and/or absorb a neutron to create the selected radioisotopeproduct. The target material may include, but is not limited to, auranium-based material, a plutonium-based material, or a thorium-basedmaterial. For instance, a target material may contain ²³⁵U. In anotherinstance, the target material may contain ²³⁹Pu. Further, it should berecognized that the target material need not be fissile directly uponfabrication, but rather could be or include a fertile material thatcould be converted into a fissile material through neutron absorption.For example, the target may include any known nuclear fertile materialwhich can be bred up through neutron absorption to the selected productand/or bred up and then fissioned to create the selected radioisotopicproduct. Fissionable material includes any nuclide capable of undergoingfission when exposed to low-energy thermal neutrons or high-energyneutrons. Furthermore, for the purposes of this disclosure, fissionablematerial includes any fissile material, any fertile material orcombination of fissile and fertile materials.

The identified target material may not be fertile or fissile. Forexample, ²³²Th may be used as a target material, which may be exposed toneutrons to yield the isotopes ²²⁵Ac or ²²⁷Ac. The isotope ²²⁶Ra isanother example, which when exposed to protons may also generate ²²⁵Ac.Yet another example is using ¹⁵³Eu as a target material, which whenexposed to fast neutron (i.e., kinetic energy above 1 keV) radiationyields ¹⁵³Sm. A further example includes using ¹⁴NH₃ as a targetmaterial, which when exposed to gamma rays may undergo a photonuclearreaction to generate ¹³NH₃.

The target material may include one or more metallic target materials,such as, but not limited to, a substantially pure metal target material,a metal alloy target material, or an intermetallic target material. Forexample, a pure metal target material may include, but is not limitedto, ²³³U, ²³⁵U, ²³⁹Pu, and/or ²³²Th. In another example, a metal alloytarget material may include, but is not limited to, uranium-zirconium,uranium-plutonium-zirconium, uranium-zirconium-hydride,thorium-aluminum, or uranium-aluminum. By way of a further example, anintermetallic target material may include, but is not limited to, UFe₂or UNi₂. It should be recognized that the above list of suitablemetallic target materials for inclusion in a target is not exhaustiveand should not be interpreted as a limitation but rather merely asexamples.

In another embodiment, the target material of a source may include oneor more ceramic target materials, such as, but not limited to, an oxidetarget material, a nitride target material, or a carbide targetmaterial. For example, an oxide-based nuclear material may include, butis not limited to, uranium dioxide (UO₂), plutonium dioxide (PuO₂), orthorium dioxide (ThO₂). Moreover, an oxide-based target material mayinclude a mixed oxide target material, such as, but not limited to, amixture of PuO₂ and depleted or natural UO₂. In another example, anitride-based target material may include, but is not limited to,uranium-nitride or plutonium-nitride. By way of a further example, acarbide-based target material may include, but is not limited to,uranium carbide. It should be recognized that the above list of suitableceramic target material materials for inclusion in the target materialshould not be interpreted as a limitation but rather merely as anillustration.

In an embodiment, the target material identification operation 304includes the determination of the complete compound or combination ofcompounds for the source material. It should be recognized that, inaddition to the fissionable materials described above, the sourcematerial may also include ancillary material, which in some cases mayinclude portions of non-fissionable material, such as, but not limitedto, radiation-inert material, neutron moderating material or neutronreflective material. Such non-fissionable material may be provided toadd strength, form, structure, or other properties to the target thatcould not be easily achieved using fissionable material alone.

It should also be noted that, in an alternative embodiment of thetargetry coupled separation method (not shown), the targetidentification operation 304 may precede the radioisotope selection.This embodiment may occur in situations where the target material isprovided and not substitutable. In this embodiment, the owner of thetarget material may wish to use targetry coupled separation on theprovided target material in order to extract some valuable radioisotopesfrom the target material in lieu of or prior to simply disposing of thetarget material.

For any given solid selected target material, a recoil distance of theselected radioisotope(s) may be determined in a recoil distancedetermination operation 306. When a nuclear reaction occurs that covertsan atom of fissile material into a radioisotope atom, kinetic energy isimparted to the radioisotope atom. The amount of kinetic energy impartedvaries based on the initial kinetic energy of the neutron, the atomicmass of the fissile atom, and the atomic mass of the direct productradioisotope, among other things. This kinetic energy causes theselected radioisotope(s) to recoil, i.e., move relative to the initialposition of the fissile atom undergoing the nuclear reaction in thesource material. The term recoil distance refers to the average distanceor range of distances which a specific radioisotope is expected to movebased on the imparted kinetic energy. Because many nuclear reactionshave been well characterized, the kinetic energy and/or recoil distancecan often be calculated or has been determined empirically for manygiven combinations of nuclear chemistry and neutron generator. Forexample, the recoil distance of fission products in uranium dioxide isgenerally described in S. G. Prussin et al., “Release of fissionproducts (Xe, I, Te, Cs, Mo, and Tc) from polycrystalline UO₂ ,” Journalof Nuclear Materials, Vol. 154, Issue 1 pp. 25-37 (1988), the recoil offission products in thorium metal is generally described in C. H. FoxJr. et al., “The diffusion of fission products in thorium metal,”Journal of Nuclear Materials, Vol. 62, Issue 1 pp. 17-25 (1976) and themigration of gaseous and solid fission products in a uranium-plutoniummixed oxide fuel is generally described in L. C. Michels et al.,“In-pile migration of fission product inclusions in mixed-oxide fuels,”Journal of Applied Physics, Vol. 44, Issue 3 pp. 1003-1008 (1973). Suchreferences allow one of skill in the art to estimate the recoil ofselected radioisotopes for a particular system.

The recoil distance determination operation 306 refers to calculating,estimating or otherwise identifying the expected recoil distance for theselected radioisotope within the selected target material. In anembodiment, the recoil distance determination operation 306 takes intoaccount the density of the target material, the particulars of theneutron generator and other aspects of the system design. The recoildistance may be determined empirically from prior experiments or may beestimated using known characteristics of the materials and atomsinvolved, such as the atom number of the direct irradiation products.The range of any particle in material can be found with the stoppingpower, which is the relationship between a particle's kinetic energy andthe range in material. For the production of radioisotope products, theenergy can be due to the recoil from the decay route, as in fission oralpha decay, or the nuclear reaction, as in fast neutron or acceleratedparticle bombardment. The product isotope's energy will need to bedetermined based on its production route. A number of routes are knownand data are available to assess the distance an energetic particle cantravel through material. The Bethe-Bloch formula provides the energyloss of particles traveling though material in units of energy distancesquared per unit mass, an example is MeV cm² g⁻¹. Stopping power andrange tables are available from numerous references, e.g., from theInternational Atomic energy Agency and the National Institute ofStandards and Technology, that can provide data to assess the recoilrange, including continuous-slowing-down approximation, of producedisotopes. The units for ranges and stopping power can be the same as theBethe-Bloch formula, or as a range in mass per area, such as g cm⁻².Programs are also available that provide ranges and stopping powers forions in materials (see, e.g., the SRIM software package available fromDr. James F. Ziegler). Once ranges or stopping powers are obtained, thedistance a particle will travel in material can be estimated using thematerial density and the particle energy. If the data for a specificproduct or nuclide cannot be found, relationships between energy loss,velocity, and charge can be used.

The recoil distance is then used in a source manufacture operation 308in order to design and create a source that, for the particularcombination of selected target and source material, preferentiallyresults in radioisotopes distributed within the source material afterthe reaction so that the radioisotopes are more readily available to theextraction material than would occur in a bulk solid, or non-poroussource. Specifically, the solid portions, i.e., the pore walls, of aporous source material (such as foams, particles, and the like) may besized to be substantially similar to the recoil distance of the selectedradioisotope product. In this manner, the anticipated recoil of theselected product can be used to improve placement of the product near anavailable surface of the source material to improve extraction of theproduct from the source (e.g., dissolution and extraction of the productwithout dissolution of the target). The term ‘available surface’ is usedto describe a location, on or near a surface of a solid source material,from which the extraction material can obtain the product. In cases suchas a source material formed as a foam or other porous structure (e.g.,manufactured pores), the structure of the source material forming thepores (e.g., the walls of the pores) may be selected and formed to havea thickness substantially similar to the recoil distance of the selectedradioactive product. In cases such as particles, one-half of theparticle size or the particle radius may be sized to be substantiallysimilar to the recoil distance of the selected radioisotope. In anotherexample, in an embodiment in which the extraction material is a liquid,the available surface of the source material is a surface that theliquid can access during the separation process without having to alterthe physical properties of the target material. In some cases, theavailable surface may include locations that are not physically on asurface of the source material, but that are close enough to anaccessible surface that the extraction material can still obtain theproduct atoms, such as through diffusion. Thus, targetry coupledseparation exploits the recoil from the nuclear reaction used to producethe selected radioisotope to simultaneously make that radioisotope moreeasily recoverable in the separation operation.

In the source manufacture operation 308, the selected source material isformed into a source based on the recoil distance of the selectedradioisotope. For example, in an embodiment the source material informed into solid grains and the size of the grains is selected based onthe recoil distance of the desired radioisotope. As a further example,if ⁹⁹Mo is the selected radioisotope product (with an anticipated decayto ^(99m)Tc, the desired product) and the selected target is an oxide of²³⁵U, then in an embodiment a source material including grains having anaverage particle size (such as diameter or average width) of equal to orless than two times (2×) the recoil distance of the ⁹⁹Mo product butgreater than 10% of the recoil distance. In another embodiment, theaverage particle size may be selected to be within ±50% of the recoildistance (0.5-1.5×) of the ⁹⁹Mo product and, in yet another embodiment,the average particle size may be selected to be ±50% of half(0.25-0.75×) the recoil distance of the ⁹⁹Mo product. In anotherembodiment, the average particle size may be selected to be within ±50%of the twice the recoil distance (1-3× the recoil distance of theselected radioisotope). In situations where there are more than oneselected radioisotope each with a different recoil distance, the recoildistance used for sizing may be selected from that of any one or theselected radioisotopes, an average of the recoil distances of some orall of the selected radioisotopes, or a weighted average based on theexpected yield of the selected radioisotopes.

In an alternative embodiment, grain size of less than 20 micrometers maybe used. In yet another embodiment, a grain size between about 0.1 to 10micrometers may be used. Generally fission products have a recoil rangeof around 10 microns in UO₂.

For solid source embodiments, the processing of the grains of sourcematerial into a solid, porous solid may include any suitable processingtechnique including one or more of sintering, milling, sieving, 3Dprinting, crystallizing, precipitating, or heating the grains of targetmaterial. The solid source may take any high surface area form such asan open-cell foam, an open lattice, an open framework, a ceramic, acloth, a thin film, a monolayer, a sponge, a nanocage, or a nanocrystal.

The nuclear reactions can also induce chemical changes that can be usedfor selective separation. Such induced chemical changes are called hotatom chemistry and described in the literature. In hot atom chemistry,the nuclear reaction changes the chemical form of the reaction productcompared to that of the target. The difference in chemistry between thetarget and reaction product, and the morphology of target, permit aseparation of the reaction product without target destruction. As anexample, a target could be a compound in a high oxidation state. Uponreaction with a neutron, the new isotope undergoes reduction and hasdifferent chemical properties than the target even though it is the sameelement as the target. The target morphology permits a separation of theproduct with the lower oxidation state without the need to dissolve thetarget. Additional detail regarding embodiments of targets, sourcematerials, and the source manufacture operation 308 are discussed withreference to FIG. 5, below.

The source manufacture operation 308 may include further selecting,creating and/or providing a suitable container for the source material.For example, in an embodiment in which neutrons are the form ofradiation used, the container may be made of a neutronically-translucentmaterial, so that neutrons are capable of passing through the container.A container may be in any suitable shape and form and may be providedwith one or more valves for allowing the easy introduction and/orremoval of the extraction material.

In the embodiment shown in FIG. 3, after the source or sources have beencreated, the sources are exposed to neutrons for some irradiation periodin an irradiation operation 310. This operation 310 may includetransporting the source(s) to the irradiation facility/equipment forsafe irradiation, for example by conveyor belt as described above. Inthe irradiation operation 310, source material is exposed to neutrons,thereby causing at least some atoms of the source material to undergonuclear fission or neutron capture to create atoms of the selectedradioisotope. This results in an irradiated source material thatcontains some amount of the selected radioisotope product within areduced amount of unreacted target as discussed with reference to FIG.11. In addition, because of the recoil from the fission reaction, atleast some of the newly created atoms of the selected radioisotope movethe recoil distance relative to the remaining, unreacted target withinthe source material. As described above, the recoil of the selectedradioisotope product may make that radioisotope more available to theextraction material such as by making the radioisotope product closer toan available surface of the source material, which may then improveextraction by the extraction material.

After the irradiation period, a separation operation 312 is performed,extracting atoms of the desired product or products from the sourcematerial. As mentioned above, the desired product may be the selectedradioisotope, a decay daughter of the selected radioisotope, or, as isthe case with ⁹⁹Mo, both. This operation 312 may include transportingthe source(s) to a separation facility/equipment, for example byconveyor system as described above. The operation 312 may also includeintroducing a storage or holding period before the separation to allowtime for decay to occur. In an embodiment of the separation operation312, the target in the source material is exposed to an extractionmaterial such as a solvent that preferentially extracts the desiredproduct from the source without substantially dissolving the remainingtarget in the source material. Embodiments of separation techniques arefurther discussed elsewhere in this disclosure, particularly withreference to FIGS. 1 and 7. In an embodiment, the remaining, unreactedsource material is not chemically reactive with or affected by theextraction solvent. Specifically, it is not necessary to dissolve thetarget to recover some of desired product from the target. Thus, thetarget is not substantially dissolved nor is its physical phase alteredby the separation operation 312. For example, in one embodiment thetarget is in a solid phase and remains in the solid phase throughout theirradiation and separation operations.

In embodiments in which sources include a container, the source materialmay or may not be removed from the container during the separationoperation 312. For example, in an embodiment, a source may compriseloose or packed individual loose grains of source material in aneutronically-translucent container, in which the grain size is based ona recoil distance of the selected radioisotope to be produced asdiscussed above. The source material grains may be repeatedly subjectedto successive irradiation and extraction operations without removing thegrains from the container. In this embodiment, a gaseous or liquidsolvent may be flowed through the container or the container may befilled or partially filled with solvent and left in the container forsome contact period of time after which the extraction material, nowcontaining at least some of the selected radioisotope, is removed.

In an alternate embodiment, rather than individual grains meeting somesize requirement based on the recoil distance, the sources may include asolid mass of target. As discussed in greater detail below, such a solidtarget may be made by sintering or otherwise bonding individual grains(which may be tailored similar to that described above with respect torecoil distance) together to form a larger source material mass. Such alarger mass may be porous to facilitate penetration of a solvent intothe porous mass, thereby facilitating contact with the generatedradioisotope.

The separation operation 312 may further include regeneration of thetarget to prepare it for subsequent irradiation. This may involve one ormore washing operations to remove extraction material from the sourcematerial prior to subsequent irradiation.

After a separation operation 312, the same source may be re-irradiatedto create more of the selected radioisotope allowing the irradiation andseparation operations 310, 312 to be repeated multiple times withoutsubstantially dissolving, changing the phase of, or removing any of theremaining mass of target material in the source. As discussed above,this allows the fissionable material to be more efficiently convertedinto the desired product that would be possible with a single neutronexposure.

The method 300 further includes a final processing operation 314 thatconverts the extracted radioisotope product into a final product orfinal form suitable for commercial use. The final processing operation314 includes separating the radioisotope from the extraction materialand may also include additional processes to purify the radioisotope.The radioisotope may then be further processed into a final formsuitable for transport and use as an industrial reagent or feedstock.

In an embodiment, the final processing operation 314 includesincorporating the radioisotope into a daughter isotope generator. Forexample, the method 300 may be used to manufacture ²²³Ra generators madefrom the radioisotope ²²⁷Ac, ⁶⁸Ga generators made using ⁶⁸Ge, ^(99m)Tcgenerators made from ⁹⁹Mo, and ⁸²Rb generators made from ⁸²Sr, to namebut a few. Daughter isotope generators and methods for manufacturingdaughter isotope generators from a parent radioisotope are known in theart. Any suitable method may be used.

For example, ^(99m)Tc generators may be created from ⁹⁹Mo in the form ofthe molybdate, MoO₄ ²⁻. To create the generator, the ⁹⁹Mo molybdate isadsorbed onto acid alumina (Al₂O₃) substrate and placed in a shieldedcolumn. When the ⁹⁹Mo atoms decay, they form ^(99m)Tc pertechnetate,TcO₄ ⁻, which, because of its single charge, is less tightly bound tothe alumina. Pouring normal saline solution through the column ofimmobilized ⁹⁹Mo elutes the soluble ^(99m)Tc, resulting in a salinesolution containing the ^(99m)Tc pertechnetate, with sodium as thecounterbalancing cation.

In an embodiment, the final processing operation 314 may be an automatedor semi-automated process. As described with reference to FIG. 2, in anembodiment a targetry coupled separation system incorporates equipmentnecessary to separate the radioisotope from the extraction fluid, modifythe radioisotope into the generator material necessary for use in adaughter generator (such as ⁹⁹Mo bound to a substrate suitable forcolumn chromatography), and package the material into the generator bodyin an automated or semi-automated process.

Container

FIG. 4 illustrates an example of a suitable container. The container 400includes a cylindrical body 402 defining an internal cavity, a top orlid portion 404 that, when engaged, seals the cavity, and a bottom 406which defines an interior chamber 414 that contains the source material.One or both of the top 404 and the bottom 406 may be removably attachedto the body 402 to allow the source material to be inserted into orremoved from the container 400. This may be achieved by any knownsystem, such as corresponding threaded portions, for example on the lidportion and in the cylindrical body (not shown). Alternatively, thecontainer 400 may be of a unitary construction and the source materialcharged through a sealable access port (not shown) or during theconstruction of the container. In the embodiment shown, two fluid flowvalves 408, 410 are provided, first valve 408 (which may be an outputvalue in some examples) in the top 404 and a second valve 410 (which maybe an input valve in some examples) in the bottom 406. In yet anotherembodiment, the container may not be completely sealed when the lid isengaged, for example, to allow gas to escape or to allow the containerto be immersed in the extraction material rather than having extractionmaterial injected into the container through a valve or access port.Although valves 408, 410 are shown at the top and bottom of thecontainer 400 respectively, one of skill in the art will recognize thatthe valves 408, 410 can be located in any appropriate location and/ororientation and do not necessarily have to be placed no opposing sidesof the container. Alternatively, one valve 408 or additional valves (notshown) may be used for any of input, output, redundancy, and/or safetymeasures of the extraction material and/or container.

A container may be of any shape, both externally and internally in thesource material chamber. Any number, type, and configuration of accessports, valves, shackles, connectors, contact points, or other ancillarycomponents may be used as desired. For example, in the embodiment showna diffuser 412 is provided so that the container may be easily used as afluidized bed or packed bed reactor. In the embodiment, the diffuser isin the form of a perforated plate with perforations sized such that thesource material (such as the particulate matter) is prevented or reducedfrom passing through it. Solvent introduced from the bottom valve 410,however, passes easily through the diffuser 412 allowing contact withthe source material. This is but one example of ancillary componentsthat could be provided on the container. For example, many differentfluidized bed reactor designs could be incorporated into a containerhaving additional ancillary components such as additional diffusers,manifolds, baffles for distributing solvent flow evenly, non-cylindricalinternal shape of the source material chamber/cavity 414, baffles fordirecting flow, etc.

In an embodiment in which neutrons are the radiation used in thetargetry coupled separation, the container may beneutronically-translucent as discussed above. Examples of suitableneutronically-translucent container materials include aluminum,zirconium, and molybdenum and alloys thereof as well as stainless steelalloys. Some or all of a container may be made from one or more of theseneutronically-translucent materials.

Containers may be made with an opening to facilitate the insertion andremoval of the physical form of source material to be used. For example,when one or more large masses of source material are used as discussedabove, a container may be provided with a relatively large opening thatallow for the insertion and removal of the masses. This would allowcontainers to be reused after the source material is spent.Alternatively, a container may be constructed around the source materialwith the intention that the source material be disposed of with thecontainer and no provision is made for removing the source material fromthe container once the target is sufficiently spent which may reducewaste and/or waste processing.

Source Manufacture, Recoil Movement, Surface Treatment Improvement

FIG. 5 illustrates an embodiment of a method of manufacturing aradioisotope-generating source in greater detail. As such, the method500 represents an embodiment of the source manufacture operation 308discussed above with reference to FIG. 3. In the embodiment shown inFIG. 5, the target material includes an oxide of a fissile or fertilematerial, such as thorium, uranium, or plutonium oxide. In variousembodiments, targets in targetry coupled separation may include oxideswhich may be manufactured using any appropriate method, although manypossible examples are provided below. Uranium oxides and plutoniumoxides with suitable target properties have been prepared andcharacterized. Actinide salts may be dissolved in solution andprecipitated to form solids.

For embodiments using solid sources, any morphology may be used,although higher surface area morphologies will have a better recovery ofproduct. Suitable high surface area morphologies include porous sourcesof: loose or sintered particles or powders; open-cell foams; 3D printed,milled, or crystallized open lattices or open frameworks; cloths; thinfilms and monolayers; sponges; ceramics; nanocages; and nanocrystals.Preferably, a solid source will have a surface area greater than 10 m²/gas measured by Brunauer, Emmett and Teller (BET) analysis.

For embodiments that use a liquid source, the target material may besolid, such as solid particles, suspended in a liquid, such as in acolloid suspension.

In the embodiment shown in FIG. 5, the method 500 begins with adissolved salt of a fissionable material in a solution as a startingmaterial, in providing operation 502. In an embodiment, the startingmaterial may be created by dissolving and mixing chloride or nitratesalts of the appropriate fissionable material in purified water. Inembodiments, the provided dissolved salt of the fissionable material maybe in a solution such as an acidic solution, a basic solution, anaqueous solution, and an alcohol solution

Next, a precipitant, such as sodium hydroxide, ammonium hydroxide,and/or oxalic acid, is mixed into the solution in a precipitant additionoperation 504. The solution is maintained at the proper conditions forthe precipitation to occur and the precipitate, an oxide of thefissionable material, is collected in a collection operation 506.Variations in the precipitation can include addition of ammoniumhydroxide, peroxide, carbonate, or oxalate. Precipitation has been usedto produce thorium, uranium, and plutonium containing oxides and isappropriate for other metal oxide formation. Any suitable method forprecipitating a fissionable material oxide, now know or later developed,may be used.

The precipitated oxide is then formed into grains in a grain formingoperation 508. This may include milling, calcining, or sintering theprecipitated oxide to form powders and/or pellets and/or any othersuitable form of the target. For example, in an embodiment of the grainforming operation 508, the precipitate may be washed with acetone andpurified water after collection, milled, and dried at 90° C. The driedprecipitate can be milled again and redried. It can be again milledand/or then calcined up to 750° C. for 1 hour. The calcined powder canbe milled and additionally or alternatively then cold pressed intopellets (of any appropriate size as determined based desired propertiesof the source material such as the recoil distance of the selectedradioisotope) for an appropriate time (which in some cases may beapproximately 2 minutes) before being sintered. In an embodiment,sintering may be under a mixture of argon and 4% hydrogen for four hoursat 1500° C.

In an embodiment, the grain forming operation 508 may include a sizingoperation to ensure either a particle size distribution of the grainsand/or that the grains have a particle size less than some thresholdsize, such as a recoil distance. Sizing of grains to obtain a desiredresult is known in the art and any suitable method of sizing grains maybe used, such as mechanical screening, filtration, and classification,electrical methods such as electrophoresis and electrostaticprecipitation, and flotation. For example, suitable equipment for grainsizing, depending on the embodiment, may include sieves; gas or liquidelutriation columns; stationary screens; grizzlies; gyrating screens;vibrating screens; centrifugal sifters; cake filters; clarifyingfilters; classifiers; and crossflow filters. In some embodiments, afterthe precipitates are formed and sized, calcination of the product yieldscompounds suitable for sintering. Sintering time, temperature,atmosphere, and oxide preparation can be varied to produce suitabletarget properties as is known in the art.

In embodiments, in addition to the sizes enumerated above with referenceto FIG. 3, grains of target may be sized to have a maximum grain size ofequal to or less than 10,000 nm (10 micrometers), or, alternatively,less than 1,000 nm, less than 100 nm, less than 50 nm, less than 10 nm,less than 5 nm or less than 2 nm. Furthermore, grains having a grainsize ranging between 1 nm and 10 mm are anticipated to be particularlyuseful based upon recoil from nuclear reactions, fission, alpha decay,or beta decay.

In an embodiment, the method 500 may be considered to include some ofthe operations of the method for selectively generating a desiredproduct using targetry coupled separation illustrated in FIG. 3. Forexample, in an embodiment of the source manufacture method 500, theselected radioisotope operation 302, target identification operation 304and the recoil distance determination operation 306 described withreference to FIG. 3 may be included in the method 500.

In the embodiment shown, the method 500 includes a source formationoperation 510 in which the grains are formed into a source material.This operation 510 is optional and not necessary in embodiments in whichloose grains are used as the form of the source material. This mayinclude combining grains into a solid mass to be incorporated into asource material, such as pelletizing the grains, making a ceramic fromthe grains, and/or making a solid matrix in which grains areincorporated.

The use of nanoparticles in the preparation of ceramics may yieldmaterials with desirable properties as a source for targetry coupledseparation. The term nanoparticle refers to grains having a grain sizeless than 100 nanometers. Compared with traditionally produced ceramics,ceramics created from nanoparticles (nanoparticle-based ceramics) havegreater hardness and higher yield strength. Nanoparticles of fissionablematerials may be produced, for example, by the precipitation methoddescribed above.

It is expected that ceramic properties derived from nanoparticles andnanoparticle synthesis routes will be useful in generating porous targetmaterial from tetravalent actinides. The following method for generatingan actinide (e.g., U, Th, and/or Pu) ceramic is proposed. First,nanoparticles of the tetravalent actinide having a selected particlesize are generated in an inert atmosphere. The particles are thensintered, for example at a temperature from 1,000-1,500° C., such as forexample 1150° C. An inert atmosphere will be maintained throughout thisprocess to prevent oxidation of the actinide metal. For example, anargon, neon, helium, nitrogen or any suitable inert gas mixture may beused. When this method was applied to zirconia nanoparticles as asurrogate for the actinide, after 2 hours of heating, a density of 93.5%theoretical was found. Density increased to 97.5% with 40 hours ofheating and reached 99% at 60 hours. The average grain size was found tobe 120 nm after 60 hours of heating at 1150° C. It is anticipated thatactinide nanoparticles will have the same or similar properties and besuitable for use in targetry coupled separation sources.

It is also expected that actinide oxide nanoparticles could be used toproduct ceramic films and membranes and that such actinide oxidenanoparticles will have desirable properties for targetry coupledseparations. The following method for generating an actinide oxideceramic is proposed. First, actinide oxide nanoparticles are created. Inan embodiment, this may be done by precipitating actinide oxide from abasic solution. In an alternative embodiment, actinide oxide particlesmay be synthesized by dissolving an actinide oxycarbonate in acidicnitric solutions, followed by hydrolysis and condensation of polynuclearactinide cations which should promote the formation of nano-sized,polymeric, oxy-hydroxide particles. When applied to zircon as surrogate,the zirconia particles so produced were found to be in the 3-6 nm range.

Creation of actinide oxide nanoparticles may also be enhanced throughvarious techniques. In an embodiment, alcohol may also be used as asolvent to generate actinide oxide nanoparticles. The alcohol solventmay induce a faster particle formation rate and produced submicrometermicrospheres due to the low solubility of hydroxide species in alcoholsolution. In yet another embodiment, adding polyethyleneimine and2,3-dihydroxybenzoic acid in the precipitation phase may to produceparticles with a suitable particle size distribution. In yet anotherembodiment, oxalate precipitation may result nanoparticles within asuitable particle size.

A hydrothermal technique may also be suitable for the synthesis ofactinide oxide nanoparticles which may be a suitable form for sourcematerial in targetry coupled separation. Urea can be used in thesynthesis of the nanoparticles produced by hydrothermal conditions. Thegeneral method for this technique is actinide precipitation under basicconditions at temperatures above 100° C. in a pressure vessel.Monoclinic nanocrystal nucleation and growth is expected to occur at1200° C. from powders produced by forced hydrolysis. The particles soproduced may then be sintered into a ceramic as described above.

In yet another embodiment, near-critical water may be used to formactinide nanocrystalline materials. An aqueous mixture of actinide isbrought to near-critical conditions and shock waves are produced bynozzle cavitation to generate actinide oxide particles. Near-criticalwater has been shown to rapidly hydrolyze and subsequently dehydratecerium and zirconium salts to form mixed ceria-zirconia nanocrystallinematerials. Shock waves produced by nozzle cavitation resulted innano-sized particles of TiO₂ and ZrO₂.

In yet another embodiment, an emulsion-combustion method may be used togenerate actinide oxide particles. In this embodiment, actinide ions inan aqueous phase are mixed into a second, flammable phase to form anemulsion. The emulsion may then be burned which will result in theactinide ions being rapidly oxidized. In the emulsion-combustion method,zirconium ions in flammable solution were rapidly oxidized uponcombustion. This method produced hollow, thin-walled particles ofsub-micrometer size.

In various embodiments, sources in targetry coupled separation mayinclude metal-organic frameworks (“MOFs”). MOFs include coordinationsolids formed from linking metal ions with organic ligands. The highsurface area compound can make suitable sources for the coupledproduction and separation of radionuclides. Lanthanide-based MOFs havebeen examined in more detail than actinide MOFs. Most of theactinide-based MOFs are based on the uranyl cation. Varying thecombination of ligands and synthetic conditions has generated a largenumber of solid-state compounds. Molecular templates have been appliedto uranyl MOFs.

In various embodiments, source materials in targetry coupled separationmay include monolayers and aggregates. Photochemical reduction ofactinides in organic solvent has been used to produce actinidemonolayers and aggregates such as particles of tetravalent uraniumphosphates. The product morphology can be varied through treatment toachieve the desired grain size or characteristic lengths for frameworksother than grains. In one monolayer example, a uranium monolayer isformed through the interaction of a pulsed laser with uranyl in atributyl phosphate organic phase. The monolayer presents as a distinctlydifferent color and can be isolated from the organic phase. Aggregationoccurs upon treatment with methanol. The uranium product properties canbe tuned through coupling photoreduction parameters and monolayertreatment.

In various embodiments, a target in targetry coupled separation mayinclude uranium metal. High surface area uranium metal material can beprepared and used as targets, with or without a container, for specificradioisotope production. In an embodiment, a uranium metal ingot can beused as a starting form for uranium metal grain formation through arcmelting. The arc melting parameters can be adjusted to produce metalgrains with desirable properties such as a grain size selected based onthe recoil distance of the selected radioisotope. Uranium metal can alsoform a high surface area structure through a hydriding-dehydridingprocess.

In an embodiment, the Kroll process may be modified to produce a highsurface area structure in the form of a porous actinide metal spongefrom an actinide tetrachloride. The Kroll process involves the reductionof uranium chloride by liquid magnesium or sodium. Electrochemicalreduction can also yield uranium metal which may have desirable targetproperties.

After the formation operation 510, the source material may be placed ina container in a packaging operation 512. Packaging of fissionablematerial into containers has been discussed above. Containers have beendiscussed in detail above with reference to FIG. 4.

Alternative Characterization of Grains

FIGS. 6A through 6C illustrate a more detailed means for characterizinggrain size than the typical approximation of a grain as a sphericalparticle with a characteristic diameter. The characteristic length 106along at least one dimension of one or more grains 104 may include acharacteristic length 106 along all dimensions of one or more grains 104of the source material 100. For example, the grains 104 of the sourcematerial 100 may be engineered such that the “height”, represented by“a,” and “width,” represented by “b” are similar in size. Therefore,notwithstanding of factors (e.g., stress or thermal gradients), aradiation product subject to recoil upon creation may efficientlydiffuse from the grain interior 110 to the grain boundary 112 along alldirections within the grain. In this context, a grain structure may becharacterized by the “grain size” of the grains 106 of the sourcematerial 100. The “grain size” may be selected such that the grains aresmall enough to allow for adequate diffusion from the interiors 110 ofthe one or more grains 104 to the boundaries 112 between the one or moregrains 104.

As shown in FIG. 6B, the characteristic length 106 along at least onedimension of one or more grains 104 may include a characteristic length106 along a selected dimension of one or more grains 104. For example,as shown in FIG. 6B, the grains 104 within the source material 100 maybe engineered to have a selected characteristic length 106 along a givendimension of the grains 104. For instance, in a grain 104 having anelongated grain structure, the grain may have a selected characteristiclength along the “thin” dimension, shown as dimension “a” in FIG. 6B, ofthe grain 104. In another instance, in a grain 104 having an elongatedgrain structure, the grain 104 may have a selected characteristic lengthalong the “thick” dimension, shown as dimension “b” in FIG. 6B, of thegrain 104. It should be recognized that the grain 104 need only have atleast one characteristic length 106 smaller than the distance requiredfor adequate diffusion due to recoil from the interiors 110 of the oneor more grains 104 to the boundaries 112 of the one or more grains 104.It is further recognized, however, that all dimensions of a grain 104may have a characteristic length 106 smaller than or equal to a distancerequired for adequate diffusion of fission product 108 from theinteriors 110 of the one or more grains 104 to the boundaries 112 of theone or more grains 104.

As shown in FIG. 6C, the characteristic length 106 along at least onedimension of one or more grains 104 may include a characteristic length106 along a selected direction 134. For example, the grains 106 withinthe source material 100 may be engineered to have a selectedcharacteristic length 106 along a given direction in the source material100. For instance, a grain 104 having an elongated grain structure mayhave a selected characteristic length 106 along a selected direction 134within the nuclear fuel. It should be recognized that engineering thegrain structures to have a characteristic length 106 along a selecteddirection 134 smaller than the length required for adequate movement ofa radiation product due to recoil from a grain interior 110 to agrain-boundary 112 may supply a more efficient means for transferringfission product such as a radioisotope product from the grain interior110.

In another embodiment, one or more grains 104 may have a characteristiclength 104 along a dimension of the one or more grains selected tomaximize heat transfer from a grain-interior 110 to a grain-boundary112. For example, the one or more grains 104 may be oriented such thattheir narrow dimensions, shown as “a” in FIG. 6C, are alignedsubstantially perpendicular to a thermal gradient 136 in the sourcematerial 100. Such an arrangement aids in the heat transfer from thegrain-interior 110 to the grain-boundary, aiding in the diffusion of afission product 108 from the grain interior 110 to its grain boundary112. By way of another example (not shown), in a cylindrical pelletfabricated utilizing the source material 100 the grains 104 of thesource material 100 may be arranged (i.e., on average the grains of thematerial may be arranged) to have their the narrow dimensionsubstantially perpendicular to the radial thermal gradient of thecylindrical pellet. It should be noted that the illustrations in FIGS.6C, 6B, and 6A represent simplified conceptual illustrations of aplurality of grains 106 should not be interpreted as schematic innature. Further, it should be recognized by those skilled in the artthat a variety of materials processing techniques (e.g., cold-workingand/or annealing, compression, or extrusion) may be implemented in orderto develop the symmetrical grain structure in FIG. 6A, and the deformedelongated grain structure illustrated in FIGS. 6B and 6C. A variety ofmaterials processing techniques are discussed further herein.

In another embodiment, the grains 104 of the source material 100 mayhave an average characteristic length 106 along at least one dimensionsmaller than or equal to a selected distance necessary for adequatediffusion of a fission product. For example, the grains 106 of thesource material 100 may have an average characteristic length along aselected dimension or direction of the grains 104 of the nuclear fuel.It is recognized that there may exist a maximum average grain size whichwill provide adequate diffusion of fission products from the interiors110 of the grains 104 to the grain boundaries 112 of the grains 104.

In another embodiment, the grains 104 of the source material may have aselected statistical distribution of characteristic lengths. Forexample, the grains 104 of the source material 100 may have a grain sizedistribution having a selected percentage of the grains having a grainsize below a selected distance. For instance, the source material 100may have a grain size distribution such that 75% of the grains have agrain size equal to or less than 5 μm, with an average grain size of 3μm. In another embodiment, the grains 104 of the source material 100 mayhave multiple statistical distributions of characteristic lengths. Forinstance, the source material 100 may have a grain size 106 distributionsuch that 25% of the grains have a grain size equal to or less than 10μm, 25% of the grains have a grain size 106 equal to or less than 5 μm,and 10% of the grains are below 1 μm. In another instance, the sourcematerial 100 may have a grain size 106 distribution such that 25% of thegrains have a grain size 106 equal to or less than 10 μm and 25% of thegrains have a grain size equal to or greater than 50 μm. In anotherinstance, the source material 100 may have a grain size distributionsuch that 25% of the grains have a grain size between 1 μm and 5 μm, 50%of the grains have a grain size between 5 μm and 10 μm, and 25% of thegrains have a grain size 106 greater than 10 μm. Applicant's co-pendingU.S. patent application Ser. No. 13/066,253, filed Apr. 8, 2011, titledNuclear Fuel and Method of Fabricating the Same, which is herebyincorporated herein by reference, includes embodiments of nuclear fuelmanufacture that could be used to create suitable target material foruse in targetry coupled separations.

Liquid Source

Regarding various other embodiments of sources, a liquid source materialcan be employed and may be coupled with continuous separation to provideradioactive isotopes. As noted above target destruction can be reducedby limiting the phase change of the target (e.g., a liquid target withor without solid or other phase ancillary materials or a suspension ofsolid target in a liquid phase source material) through separation. Inthis embodiment, a liquid source can be a molten salt or solution phase.The liquid source can flow through an irradiation location or may becontained in a container that is passed through the irradiationlocation. The resulting radionuclides produced from a liquid source canbe separated, isolated, and purified from the target using conditionsand automation procedures as described for the solid source. Suchseparation may use a liquid-liquid extraction process, a liquid-gasextraction process, an electrochemical process, or, alternatively, aliquid-solid extraction process such as passing the irradiated liquidsource over a solid material adapted to remove the desired product(s)from the liquid phase. For example, in a liquid-liquid extractionembodiment, under certain conditions the target may be immiscible in orotherwise separable from the extraction material to facilitateseparation of a liquid extraction material from the liquid sourcematerial after a sufficient contact time. A liquid source embodiment mayhave similar benefits from target reuse and waste reduction, but thesource configuration and flow may entail additional considerations thansources of solid material.

A liquid fuel recycle system for removing fission products fromsalt-based fuels and recycling the fuels back to the reactor may bechemically similar to the process developed for metallic fuels.Supercritical CO₂ separation, in particular, takes advantage of theproperties of the salts, which are, by themselves, insoluble in sCO₂.Extractants, such as diketones, may be used to draw select metals intothe sCO₂ phase as described herein. Physically, the liquid fuel recyclesystem may be made to avoid pressurization of the reactor vessel duringa leak in the sCO₂ system. Additionally, the salts in their liquidstates may be at temperatures high enough to dissociate or degrade thediketones. To avoid both of these obstacles, a liquid fuel recyclesystem may be designed such that the molten-salt is pumped external tothe reactor vessel and injected into a vessel containing the sCO₂. AsCO₂ system may be maintained at a temperature low enough to solidifythe molten-salt, resulting in a high surface area solid. Provided thesCO₂ can be maintained at a sufficiently low temperature, thebeta-diketones or other appropriate extractant(s) may be co-mixed withthe sCO₂ during salt injection, avoiding dissociation.

Alternatively, the extractant may be injected into an extraction vesselin a batch-wise fashion following salt injection. In either case, theresult is a salt solution of (selected) metal-complexes solvated in thesCO₂ diketone solution. The salt solution may then be pumped to asecondary system where temperature or pressure is adjusted to remove themetal complexes (product) from the salt solution without substantialdestruction of the target in the molten salt fuel. Again, it is likelythat the metal complex is removable form the salt solution withoutdropping the CO₂ to a gaseous state (below the critical point) viaheating, cooling, or both. Heat may be used to volatilize the metalcomplexes so that a separate gas phase occurs within the sCO₂ solution.The sCO₂ may alternatively be cooled or heated near and above thecritical point where its solubility typically changes significantly withchanges in temperature and pressure, resulting in a separate,liquid-metal complex phase which was forced out of solution due tochanges in thermodynamic condition. This phase can then be transferred,such as by way of pumping, from the extraction system to a systemdesigned for interim or long term storage. Whether further heating orcooling is used to separate the metal complex or other product,ultimately further heating can be used to thermally decompose thediketones, leaving behind the metal fission product(s).

Separations of Radioisotope(s) from a Source

Embodiments suitable for use in one or more of the separation operationsdescribed above will now be described in greater detail. As discussedabove, embodiments of the separation of the desired product(s) from asource may include exposing at least some of the source material to anextraction material that preferentially extracts the selectedradioisotope product from the source material without removingsubstantially any of the target or requiring the target to be dissolvedor to otherwise require a change in the phase or physical form of thetarget. This allows the target to be reused in a subsequent neutronbombardment with little or no regeneration or post-separationprocessing.

In an embodiment, the separation process generally involves thepreferential isolation of the desired product(s) created by neutronbombardment from a solid phase source material. The separation isperformed without dissolution of the source material or the targetwithin the source material using a solvent as the extraction material.As mentioned above, the targetry coupled separation can exploit therecoil from the nuclear reaction used to selectively tailor and producethe target nuclei and target material to make the desired product moreeasily recoverable in the separation operation. Additionally and/oralternatively to the recoil, chemical differences between the target andthe product nuclei can be selected, tailored and/or exploited to achievea preferential separation of the desired product(s). Additional stepsmay be desired to remove the extracted radioisotope product from theextraction material and, in subsequent steps, further purify the desiredproduct. Additional purification may utilize any one or more appropriatetechniques as are known in the art, including column chromatography,precipitation, electrochemistry, ion exchange, sorption, filtration, andsolvent extraction.

Based on the source material composition, properties, and/or morphology,in various embodiments, the nuclear reaction may separate the productnuclei from the source or may physically move the product nuclei near oronto an available surface of the source material, which causes it to bemore accessible to an extraction material, or may induce a chemicalchange that can be utilized to achieve separation. Appropriateextraction material can be selected, formed, introduced and/or activatedin the separation process to exploit differences in the extractionproduct and the target, with the extraction product being either thedirect (selected) product of the neutron bombardment or an indirect(decay daughter) product of the selected radioisotope. The desiredproduct is amenable to separation due to the behavior of its chemicalform in a solid, liquid, or gas phase. Additionally, the chemicalprocesses do not appreciably dissolve the source or at leastsubstantially reduce dissolution of the target, thereby leaving thetarget in a state to be reformed for further irradiation.

Various separation treatment options are available. In variousembodiments the target may be removed from the irradiation generator andtreated. The treatment can use any single and/or combination of anyappropriate process including chemical, electrochemical, thermal,filtration, pressure, fluidized bed, and gas phase methods. Solutionphases can include any one or more phases including aqueous phases,organic phase, ionic liquids, molten salts, suspensions, andsupercritical fluids. The chemical composition of the gas phase can alsobe varied in composition of gases, temperature, flow rates, pressure,etc.

Illustrative separation processes and methods can include any one ormore of members of the group comprising extraction, liquidchromatography, gas chromatography, capillary chromatography,crystallization, precipitation, filtration, distillation, fractionaldistillation, electrophoresis, capillary electrophoresis, magneticseparation, evaporation, flotation, cloud point, micellar, flocculation,electrochemical methods, volatilization, and sublimation. The separationprocess can be performed in the source's container, thereby precludingremoving the target from the irradiation container. Alternatively, ifthere is no container, the source material may be placing inside achemical reactor or other container and then removed for subsequentre-irradiation after the separation is complete. If desired, theseparation process can utilize an automated chemistry system, such asthose produced by Chemspeed Technologies, Skalar, Human Diagnostics,Randox or any other appropriate automated chemistry system.

FIG. 7 illustrates an embodiment of general separation method suitablefor use with targetry coupled separation. The method 700 obtains adesired product or products from a source material that has previouslybeen irradiated so that at least some of the desired product isdistributed throughout the source material. In the embodiment shown,such a source material is provided in operation 702.

In a selection operation 704, an extraction material that removes thedesired product or products from the source material withoutsubstantially dissolving the source material is selected and preparedbased on the desired product to be removed and the characteristics ofthe source material. For example, in an embodiment, the extractionmaterial may be a solvent that dissolves the desired product but doesnot dissolve fissionable material in the source material. In yet anotherembodiment, the extraction material may be a solvent containing anextractant, such as a ligand, that will bind to the desired product(thereby making it soluble with respect to the extraction material) butwill not bind to fissionable material. If there are multiple desiredproducts, one extractant may be suitable or, alternatively, multipleextractants may be selected. Such a ligand should be soluble in thesolvent under temperature and pressure conditions of the contactingoperation.

In yet another embodiment, and as will be discussed in greater detailbelow, the solvent may be sCO₂ and the selected ligand or ligands form acarbon dioxide soluble chelate with the radioisotope. Again, such aligand should be soluble in the solvent under temperature and pressureconditions of the contacting operation. For example, for removal usingsCO₂, the ligand concentration may be up to 0.5 mole/liter and thetemperature and contacting time may be varied. However, sufficientremoval is anticipated to occur at temperatures below 220° C. at 1 atmwith a contacting time of 30 minutes or less. Examples of possibleligands include a fluorinated β-diketone and a trialkyl phosphate, or afluorinated β-diketone and a trialkylphosphine oxide. Further examplesinclude dithiocarbamates, thiocarbazones, -diketones and crown ethers.Inorganic ligands, including nitrates, sulfates, thiocynates, cyanates,and other similar compounds may also be used. A ligand may be providedwith one or more functional groups selected to enhance the ligandsability to bind and remove desired products. Such functional groupsinclude hydroxyl, carbonyl, diketones, aldehyde, haloformyl, carbonateester, carboxylate, ester, ether, peroxy, amine, carboxamide, imide,imine, nitrate, cyanate, thiol, sulfide, sulfinyl, sulfonyl,thiocyanate, isothiocyanate, phosphate, and phosphono groups.

Next, the source material is exposed to the extraction material, in acontacting operation 706 which in some cases may include adding theextraction material to the source material. Various actions may beperformed to enhance the contact between the extraction material and thesource material, again depending on the characteristics of thecomponents involved. For example, if the source material is solid, thecontacting operation 706 may include contacting the source material witha liquid extraction material for a residence time. As a result, anextraction material and radioisotope liquid mixture is created as thedesired radioisotope product is dissolved from the source material.Alternatively, if the source material is a liquid, the contactingoperation 706 may include contacting the source material with animmiscible liquid extraction material for a residence time. This resultsin a two-phase liquid mixture containing a first phase of bulk materialand a second phase of extraction material with the dissolved desiredproduct.

The contacting operation 706 may also include other actions to assist inseparation. For example, in an embodiment the contacting operation 706includes agitating one or both of the source material and the extractionmaterial during at least a portion of the residence time. In yet anotherembodiment, the contacting operation 706 includes changing a temperatureof one or both of the source material and the extraction material duringat least a portion of the residence time. And, in yet anotherembodiment, the contacting operation 706 includes changing a pressure ofone or both of the source material and the extraction material during atleast a portion of the residence time.

In yet another embodiment in which the source material is in the form ofsolid grains stored in a container, the contacting operation 706includes inserting an amount of the extraction material into thecontainer and retaining the extraction material in the container forsome predetermined residence time.

After a selected residence time, the extraction material, now includingthe dissolved desired radioisotope product or products, is removed fromcontact with the source material in a removing operation 708. This mayinvolve simply draining liquid extraction material from the sourcematerial or may require more active processing such as using centrifugalforce, heating, cooling, pressurizing, or depressurizing to remove theextraction material.

The desired product or products may then be separated from theextraction material and converted into a final product in a separationoperation 710, substantially as described above with reference to finalprocessing operation 314 of FIG. 3.

Volatility-Based Separations

It will also be appreciated that in various embodiments of an extractionprocess may include a rapid, volatility-based separation that can beused to isolate the desired products from irradiated sources.Volatility-based separation embodiments can exploit the formation ofhalides (F⁻, Cl⁻, Br⁻, I⁻), carbonyl (CO), and diketone based ligandssuch as hexafluoroacetylacetonate (“hfac”) (FIG. 9) to produce volatilemetal compounds. The formation of volatile fluorides with nuclearmaterials is known. Chlorination has also been examined, and foundsimilar to fluoride behavior. The existing differences can be exploitedand extended to the other halides for tunable separations. This isreadily performed in the Van Arkel process for obtaining pure Zr fromZrI₄. Carbonyls are used in the Mond process to form volatile Nispecies. The fission products Mo, Tc, Ru, and Rh also form carbonylspecies, with Mo(CO)₆ being a primary example of a volatile product. Thehfac complexes are known to be volatile for a range of elements. Thiscan provide a rapid and selective separation of radionuclides.

Formation of halide, carbonyl, or hfac complexes can be exploited forthe separation of a range of elements from starting material based ondifferences in volatility. Targeting the specific formation of volatilespecies can achieve separations that can be rapid and selective. Anadditional benefit is that volatile complexes may be used as metal vapordeposition precursors. Thus, a pure sample of the product could beobtained directly from volatile complexes formed in the reactionmixture. It is to be noted that decay of the product or product of theirradiation may require further processing for product generation (e.g.,a product precursor is the result of irradiation of the source).

In an embodiment, using the Mond process, an irradiated UO₂—containingsource material in a granular form having ⁹⁹Mo solid distributedthroughout the grains of the source material as the result of theprevious irradiation may be exposed to carbon monoxide in a vessel,container or chamber maintained at a pressure from 0.5 to 5 atm andtemperature from 50-60° C. According to the Mond process, this willconvert at least some of the ⁹⁹Mo to ⁹⁹Mo(CO)₆. Relatively more⁹⁹Mo(CO)₆ may be created by extending the exposure time and by othermethods such as agitating the source material to provide bettercontacting of the carbon monoxide gas with the surface of the sourcematerial. The boiling point of ⁹⁹Mo(CO)₆ (approximately 156° C.) issubstantially less than the melting point of UO₂ (approximately 2,865°C.). Therefore, volatilization can be easily achieved by raising thetemperature of the source material after the carbon monoxide contactingoperation to a temperature above the boiling point of ⁹⁹Mo(CO)₆.Furthermore, by keeping the temperature below the melting point of theUO₂ after the ⁹⁹Mo(CO)₆ has been driven off, the source material isunaffected and ready for a subsequent irradiation operation.

In another more generalized embodiment, an amount of irradiated sourcematerial having desired product distributed throughout the sourcematerial as the result of the previous irradiation may be exposed to aF⁻, Cl⁻, Br⁻, I⁻, CO, or diketone based ligands in a vessel, containeror chamber under conditions that cause the desired products to form avolatile compound of the desired product, but that do not alter thetarget material. Relatively more volatile desired products may becreated by extending the exposure time and by other methods such asagitating the source material to provide better contacting with thesurface of the source material. Subsequently, as long as the boilingpoint of the desired product compound is below the melting point of thetarget material, volatilization can be easily achieved by raising thetemperature. The remaining source material is unaffected and ready for asubsequent irradiation operation.

Supercritical Carbon Dioxide Separations

As mentioned above, another separation technology suitable for use intargetry coupled separations is supercritical carbon dioxide. The sCO₂extraction described herein may also be suitable for use in removingfission products from nuclear fuel in addition to removing desiredproducts from targetry coupled separation sources. Supercritical CO₂ hasbeen examined for extraction on metals and metalloids from both aqueousand solid solutions. Accordingly, sCO₂ combined with various ionicliquids (ILs) can be utilized as ligands to extract metal ions fromsolutions. Similar methods may be used to extract metals or metalloidsfrom solid materials, such as contaminated paper, fabrics, or evensoils. Current irradiated fissionable material recycling techniquesusing sCO₂ solutions require dissolution of the irradiated material intoa solution. Using the sCO2 separation techniques described herein, itmay be possible to treat used fuel source material (including nuclearfuels considered for molten-salt reactors) with sCO₂ in a manner whichdoes not require dissolution. As an example, metal fuel from a breed andburn reactor such as a traveling wave reactor (TWR) can be treated witha sCO₂ system that does not dissolve the U metal but does removeselected fission products (with high cross sections for parasiticabsorption). A sCO₂ system may be capable of selectively removing theseelements and their corresponding isotopes. A list of elements soluble inILs is shown in Table 2.

TABLE 2 Occurrence of selected elements in TWR spent fuel and ILsolubility. Fractional Fractional Element Absorption Element AbsorptionPd 2.38% Ru101 1.18% Ru 1.95% Pd105 1.13% Sm 1.25% Tc99 1.02% Mo 1.21%Rh103 1.02% Cs 1.16% Pd46 0.73% Tc 1.02% Cs133 0.73% Rh 1.02% Mo97 0.45%Nd 0.85% Sm149 0.43% Xe 0.41% Ru102 0.41% Eu 0.30% Mo95 0.41%

For ILs, the sCO₂ may be useful as a means of introducing uranium intothe IL. In other cases, it may be appropriate to have direct dissolutionof oxides into an IL. Metals of interest to nuclear waste processing,such as actinides, lanthanides, and transition metals, have beencharacterized chemically using highly soluble fluorinated β-diketones insCO₂. Extraction can be accomplished by using appropriate chelatingagents as extractants. For example, La and Eu extraction with greaterthan 90% effectiveness has been demonstrated using fluorinated diketonescombined with tri-butylphosphate (TBP). In this process, a roomtemperature ionic liquid, an imidazolium-based1-butyl-3-methylimidazolium (BMIM) withbis(trifluoromethylsulfonyl)-imide (also known as Tf₂N⁻, which isproperly described as (CF₃SO₂)₂N⁻) was used as a complexing agentbecause of the complexing agent's ability to solubilize CO₂. In thismanner, a full water/RTIL/sCO₂ system is developed. A similar processwith other ionic liquids and metal chelating agents (extraction agents)and is summarized in Table 3. Note that Eu and La are both extractedwith all systems except when using thenoyl tri-fluoroacetone (TTA)without TBP. The latter only extracted La while not separating(extracting) Eu. 1001721 For example, for removal using sCO₂, the ligandconcentration may be up to 0.5 mole/liter and the temperature andcontacting time may be varied. However, sufficient removal isanticipated to occur at temperatures below 220° C. at 1 atm with acontacting time of 30 minutes or less. The extractions performed inTable 3 were carried out with the extractant/sCO₂ mixture at 150 atm forone hour at 50° C. The extractions show that sCO₂ separation should besuitable for use on irradiated source material including nuclear fuel,nuclear waste material, and targetry coupled separation sources.Further, the extractions show that β-diketones can be used toselectively bind with oxides or metal in the presence of fissionablespecies such as uranium. Based on this information, it is anticipatedthat β-diketones can be used to selectively bind with radioisotopeoxides or metals while not substantially dissolving fissionable materialregardless of its origin.

TABLE 3 Degree of extraction (%) of EUIII and LaIII from BMIMTf₂N withdifferent beta-diketones (with or without TBP). Eu³⁺ La₃₊ HFA w/o TBP90.5 90.4 HFA w/TBP 99.9 92.6 TTA w/o TBP — 87.1 TTA w/TBP 95.5 90.5 HFA= hexafluoroacetylacetone, TTA =4,4,4-trifluoro-1-(2-thienyl)-1,3-butanedione

Further examples of possible ligands include dithiocarbamates,thiocarbazones, β-diketones and crown ethers. Inorganic ligands,including nitrates, sulfate, thiocynates, cyanates, and other similarcompounds may also be used. A ligand may be provided with one or morefunctional groups selected to enhance the ligands ability to bind andremove desired products. Such functional groups include hydroxyl,carbonyl, diketones, aldehyde, haloformyl, carbonate ester, carboxylate,ester, ether, peroxy, amine, carboxamide, imide, imine, nitrate,cyanate, thiol, sulfide, sulfinyl, sulfonyl, thiocyanate,isothiocyanate, phosphate, and phosphono groups.

In general, an obstacle to CO₂ solvation is low solvent power of CO₂(non-polar). Metals and metal chelates have low solubility in sCO₂ withCO₂ solubility parameters in the range of 4-5 cal/cm³. This can beovercome by adding CO₂-philic functional groups such as fluoroethers,fluoroacrylates, fluoroalkyls, silicones, and certain phosphazenes.Fluorinated beta-diketones (with and without tributyl phosphate) havebeen demonstrated in current techniques to extract a variety of metals.Bis(trifluoroethyl) dithiocarbamate exhibits higher solubility thannon-fluorinated counterparts; 10⁻⁴ mol/L for fluorinated vs. 10⁻⁶ to10⁻⁷ mol/L for non-fluorinated. As another example,Diethyldithiocarbamate (DDC) can be 3-800 times less soluble in sCO₂ at100 atm than bis(trifluoroethyl)dithiocarbamate (FDDC). Since sCO₂density change is nearly linear with pressure, the solubility alsochanges nearly linearly with solubility increasing with increasingpressure.

Lanthanides, actinides, copper, arsenic, and antimony (and otherproducts of irradiated sources) can have concentrations on the order of10⁻⁴ mol/L CO₂. Water and soil extraction has been demonstrated incurrent techniques with 1000-10000 molar ratio of chelate to metal insolution.

In large scale processes, it may be impractical to transition sCO₂ tothe gas phase and remain economical since it may require eitherrecompression of the CO₂ to the supercritical state or a steady supplyof high pressure CO₂, not to mention the safety risk inherent toconfining a high pressure solution of a highly compressible fluid.Furthermore, the off-gas CO₂ may need to be collected in a containercapable of further decontamination or disposal, due to some residualradioactive materials or decay products potentially remaining in thecarbon dioxide gas.

Some current techniques have a ‘back extraction’ process which does notrequire gasification of the sCO₂ as part of the separation of theradioisotopes from the sCO₂. In this type of process, metal or metalloidspecies are removed from solid or liquid solutions by usingsupercritical fluids to form a metal or metalloid chelate. Thesupercritical fluid will typically contain a solvent modifier, such as afew percent H₂O or MeOH. The metals or metalloids are thenback-extracted from the sCO₂ solution by using an acidic solution, onewhich is preferably halogenated. By back extracting to another (aqueous)solution, decompression of the sCO₂ is avoided. What is left is theother solution bearing the selected radioisotopes and sCO₂ that can bereadily reused. This is particularly advantageous in an automated systemand in a continuous treatment, although even in a semi-automated, batchtreatment system the ability to recycle sCO₂ without the added step ofrepressurization would be cost-advantageous. Back extraction may, or maynot remove the ligand with the radioisotope product. In an embodiment,fresh ligand may need to be added to the sCO₂ before it can be reused asan extraction material. It should be noted that ILs could also be usedfor the back extraction process.

FIG. 10 illustrates an embodiment of a method of extracting a firstradioisotope product from irradiated fissionable source material. Themethod 1000 begins with an irradiated fissionable source materialillustrated by the providing operation 1002. The irradiated fissionabletarget material may contain a plurality of radioisotopes in addition tothe desired radioisotope product. Examples of desired radioisotopeproducts include ⁹⁹Mo, ²³⁸U, ¹³¹I, ⁵¹Cr, ²²⁵Ra, and ²²⁵Ac.

Based on the desired radioisotope product or products and thecharacteristics of the target material, a ligand is selected in a ligandselection operation 1006. In an embodiment, a ligand is selected that issoluble in supercritical carbon dioxide (sCO₂), forms a chelate with thedesired product, and does not form a chelate with the target material.For example, in an embodiment, the desired radioisotope product is ⁹⁹Mo,the irradiated target material is ²³⁵U and the ligand known to complexwith molybdenum. Examples of other suitable ligands are provided above.

Next, the identified ligand is dissolved into sCO₂ to form a sCO₂-ligandsolution in an extraction material preparation operation 1006. If theselected ligand is not particularly soluble in sCO₂, this operation 1006may also include modifying the ligand to make it more soluble, such asby adding CO₂-philic functional groups such as fluoroethers,fluoroacrylates, fluoroalkyls, silicones, and certain phosphazenes. Inan embodiment, the ligand may be a fluorinated β-diketone and a trialkylphosphate, or a fluorinated β-diketone and a trialkylphosphine oxide. Inanother embodiment, the ligand may be selected from dithiocarbamates,thiocarbazones, β-diketones and crown ethers.

The sCO₂-ligand solution is then placed in contact with the irradiatedsource material for a contact time, in a contacting operation 1008. Asthe selected ligand forms a complex with the desired product, a resultof the contacting operation 1008 is a sCO₂-radioisotope complexsolution. In an embodiment, the irradiated source material is in acontainer and the contacting operation 1008 includes passing thesCO₂-ligand solution through the container.

The contacting operation 1008 may also include performing additionalactions to enhance the mass transfer of the radioisotope product intothe sCO₂-ligand extraction material. For example, in an embodiment inwhich the irradiated source material is in the form of loose or looselypacked grains in a container, the contacting operation 1008 may includepassing the sCO₂-ligand extraction material through the container,essentially using the container as a packed bed reactor by forcing thesolution through the bed of grains. In yet another embodiment, thesCO₂-ligand solution may be passed through the container at a flow ratesufficient to fluidize the plurality of grains within the container, ineffect using the container as a fluidized bed reactor. In yet anotherembodiment, the irradiated fissionable source material may be in liquidform and contacting includes agitating the fissionablematerial/sCO₂-ligand solution mixture.

After the contact time, the sCO₂-radioisotope complex extractionsolution is then removed from the irradiated source material in aremoval operation 1010. In this operation, care may be taken to preventthe fissionable target material from being removed with thesCO₂-radioisotope complex extraction material so that substantially allof the irradiated fissionable target material remains together in itsoriginal, physical form, e.g., a powder or ceramic. The reader willunderstand that a perfect system is not possible, and that some deminimis amount of irradiated material may be removed with the extractionmaterial. However, systems in which less than 1% by weight or less than0.1%, 0.01%, or 0.001% of the original amount of irradiated material isremoved with the sCO₂-radioisotope complex solution should be readilyachievable.

Next, the desired product and/or a further decay daughter product of thedesired product is separated from the sCO₂ in a separation operation1012. This may be by back extraction of the sCO2 or may involve reducingthe sCO2 to subcritical. This may include removing the ligand-productcomplex or, alternatively, may include removing only the product. In anembodiment, a back extraction is used in which a sCO₂-ligand solution isalso generated from the separation operation 1012 that is suitable forreuse without decompressing and repressurizing the sCO₂ ligand solution.In an embodiment, this may be achieved by contacting the sCO₂-productcomplex solution with an acidic solution, thereby generating anacid-product solution and a regenerated sCO₂-ligand solution.

Supercritical Carbon Dioxide Separation for Reformation of Spent Fuel

Metallic fuel, including those metal fuels appropriate for vented pinconfigurations and/or a traveling wave reactor, typically includes metalfuel capable of high burn-up contained within vented, ferriticmartensitic stainless steel cladding. At the end of life, the fuelgenerally has a highly porous matrix of metallic form fuel and solidfission products which precipitated from the fuel during the burn cycle.

FIG. 8 illustrates an embodiment of a method for the reformation ofnuclear fuel using sCO₂. Reformation of fuel after irradiation generallymay be designed to allow treatment of the entire fuel assembly forfission product, lanthanide, or actinide removal treatments withoutmodification of the nuclear fuel assembly or fuel pins contained within.Using the example of a sealed vessel with targetry coupled separations,a previously burned nuclear fuel assembly source material may be placedinto a sealable pressure vessel in a container operation 802.

The vessel is then filled with pressurized sCO₂ and one or moreextractant (such as diketones, or any other appropriate agent) to createan extraction material in the absence of an IL or aqueous component inoperation 804. Because of the presence of a vent in the existing fuelassembly for fission gas venting, and the nature of supercriticalfluids, the sCO₂-extractant solution will work to fill the fuel pin andthe matrix of porous fuel (i.e. supercritical fluids behave as lowsurface tension, low viscosity fluids which fill the volume they arecontained within). The extraction material will begin to solvatetargeted fission products (or other materials, if so desired and aproper ligand chosen), leaving the uranium metal matrix unaffected. Thefission products will then begin to diffuse out of the fuel sourcematerial such that the concentration of the overall system tends towardequilibrium.

The extraction material containing the dissolved fission products canthen be slowly released from the pressure vessel in an extractionmaterial removal operation 806. New, clean extraction material may ormay not be added to the pressure vessel during the removal operation806. Agitation, heat and/or continued pressurization anddepressurization may be applied to the system to enhance the solvationrate. For example, the system may operate at greater than 7.5 MPa(approximate critical point at 51° C.) and be oscillated by +/−0.1 MPato enhance ‘pumping’ of extraction material in and out of the porousfuel.

The extraction material removed from the system, containing the elementsand isotopes removed from the used fuel, is directed toward anothervessel in a collection operation 808.

In the separation vessel, the sCO₂ in the extraction material can bebrought to below the critical point and converted to the gaseous phasein operation 810. By reducing the CO₂ below the critical point, theextractant and the fission products are separated out of the CO₂ andcollect as a liquid phase in vessel.

Next, a volatilization operation 812 can be performed on the collectedliquid phase extractant and fission product mixture in which theextractant is brought to above its volatilization temperature andconverted to a vapor phase, leaving behind the selected element orisotopes. This may be done in the same separation vessel as thesub-critical operation 810 or the extractant-fission product liquidmixture may be moved to a different vessel for this operation.

Variations of this scheme may be used as appropriate. For example,lowering the solution to below the liquidus point of the carbon dioxidemay be preferred if the chosen extractant and liquid CO₂ are insoluble.Another alternative may be to raise the temperature of the supercriticalsolution to above the volatilization point of the extractant (e.g.greater than 100° C. to 200° C.) or to above the decompositiontemperature (e.g. greater than 200° C. to 300° C.). In either case, themetal may substantially or partially precipitate from the sCO₂ once theextractant is lost. Removal of the extractant vapor or decompositionproduct can be accomplished by a gas phase separation or, as above, byconverting the CO₂ to a liquid phase. Furthermore, the solution maychange temperature or pressure from a first supercritical condition to asecond supercritical condition, the second condition having a solubilityof the extractant lower than the solubility of the first condition. Bythis process, all or a portion of the extractant may be recoveredwithout leaving the supercritical state.

Removing fission products from the fuel assembly may greatly enhance thedisposability of the fuel assemblies, as >90% of targeted fissionproducts may be removed with >90% capable of being removed with multiplesCO₂ solution treatments. In some cases, it may be advantageous to applymultiple cycles such as repeated treatments or multiple differenttreatments, each with a different extraction material, to increase theremoval of fission products. For example, in some cases, two treatmentscould give 99% removal of accessible fission products whereas threewould give 99.9% and so forth. Any appropriate factors may be used todetermine the number and/or type of processing treatments and may bebased on fission products dissolved or stuck inside the solid fuelmatrix where sCO₂ solution cannot penetrate. It should be noted,however, that it may be possible to operate at temperature andtimescales which would allow for diffusion of solution soluble metalsout of the fuel matrix and into solution. This may lower the short termheat load of the spent fuel assembly, decrease the dangers of handlingand transporting the assembly, and make it more suitable for long-termdisposal.

An alternative to spent fuel disposal would be to re-use the fuelassembly once the fissions products are removed as a source in atargetry coupled separation method, such as described above. The fuelassembly could be transported to a targetry coupled separation facilityfor this or processed in the same facility that created the spent fuel.A fuel assembly may be used as a source without modification or it maybe processed to improve the targetry couple separation effects, such asby converting the spent fuel into grains of an appropriate size for theradioisotope products of interest to the targetry coupled separationfacility.

For example, in an embodiment the facility is a breed and burn typereactor such as a TWR. In this embodiment, the fission products may beremoved and then a thermo-mechanical treatment is performed within thepressure vessel used for solvation. The thermo-mechanical treatmentmodifies the structural material for continued in-reactor use. Toenhance the treatment, after the fission products are removed, thevessel and contained assembly may be brought to significantly highertemperatures (which could be made to exceed the fuel melting point) andpressures (10's of MPa's).

A system using targetry coupled separation may remove fission productsprior to the end of life by incorporating the separation process such assCO₂ process into the fuel management or ‘shuffling’ cycle to removefission products periodically during irradiation (operation of thereactor). For example, some TWR re-fueling systems incorporate a sealedenclosure for raising the assembly out of the vessel. In such systems,the existing enclosure also contains cooling capability to manageassembly decay heat. These systems may be made more robust such thatfission products may be removed, in containment, with minimal systemmodifications. This allows sCO₂ extraction to be done as an integralpart of the shuffling operation. Such a system would not require largevessels and piping, due to the high density of sCO₂. Concentrations ofgreater than 10-4 kg metal/kg solution are possible. At end of life,each assembly contains the maximum amount of fission products, on theorder of 50 kg. The solution density is on the order of 1000 kg/m3.Therefore only 5 m³ of sCO₂ solution would be needed in some cases tocontain all the fission products in a single assembly. Treating theassembly at more frequent intervals would obviously reduce this maximumvolume. Furthermore, since the CO₂ may be separated from the fissionproducts and re-entered into the system, the inventory can beadditionally reduced.

Irradiated Material Reprocessing

FIG. 12 illustrates an alternative embodiment of a method forselectively generating a desired radioisotope using targetry coupledseparation. The method of FIG. 12 differs from that of FIG. 3 in thatthe irradiated target material is provided as the starting material,thus limiting the options of which desired products may be selected.This may occur, for example, when a quantity of spent nuclear fuel isavailable and it is desired to use targetry coupled separation torecover some value from the spent fuel. Such an example includes theproduction of ²²³Ra from ²²³U which is a waste product from the thoriumfuel cycle.

The method 1200 begins in operation 1201 with provision of an amount ofirradiated source material, which may include some amount of both targetand ancillary material, to be used in targetry coupled separation. Theinitial source material may be spent nuclear fuel, nuclear wastecontaining some amount of fissionable material, or some other materialand may include any target material as described above.

The initial source material is then characterized to determine whatradioisotopes are within the material in a characterization operation1202. The initial source material may or may not be suitable fortargetry coupled separation without further processing and/or itsincorporation into a source material. Thus, the characterizationoperation 1202 also determines if the form of the initial material canbe modified to enhance the separation of any particular radioisotopes.

A selection operation 1204, similar to that described above withreference to FIG. 3, is then performed. In this operation 1204, however,because the initial source material is known, the range of radioisotopesthat may be selected is limited to those that can be obtained from theinitial material. In an embodiment, more than one radioisotope may beselected.

As already noted, some desirable radioisotopes may not be directproducts of an irradiation operation. In those situations, the selectionoperation 1204 may be equally considered a selection of the decay chainor a selection of any of the radioisotopes in the decay chain.

A material processing operation 1206 may then be performed. The initialsource material is processed into one or more sources. In an embodimentwhere re-irradiation is to occur, this processing may be done based onthe recoil distance of the selected radioisotope, as described withreference to FIG. 3. The processing operation 1206 may be as simple asplacing the initial source material in a container. In anotherembodiment, the initial source material may be processed, physicallyand/or chemically, to make the form of the source material more suitablefor the separation operation. For example, an initial material may becrushed and sieved to generate particulates having a selected particlesize. As mentioned above, it the initial source material is to bere-irradiated, this sizing may be done based on the recoil distance ofthe selected radioisotope. Such processing may further include sinteringthe particulate into a ceramic, as described herein. Additional detailregarding embodiments of targets, source materials, and the processingoperation 1206 are discussed with reference to FIG. 5, below.

The processing operation 1206 may include further selecting, creatingand/or providing a suitable container for the source material. In anembodiment in which re-irradiation using neutrons will occur, thecontainer may be made of a neutronically-translucent material, so thatneutrons are capable of passing through the container. If re-irradiationwill not occur, then a container of neutron-absorbing material may beselected. A container may be in any suitable shape and form and may beprovided with one or more valves for allowing the easy introductionand/or removal of the extraction material.

After the source or sources have been created, a separation operation1208 is performed, extracting atoms of the desired product or productsfrom the source material. As mentioned above, the desired product may bethe selected radioisotope, a decay daughter of the selectedradioisotope, or, as is the case with ⁹⁹Mo, both. This operation 1208may include transporting the source(s) to a separationfacility/equipment, for example by conveyor system as described above.In an embodiment of the separation operation 1208, the source materialis exposed to an extraction material such as a solvent thatpreferentially extracts the desired product from the source withoutsubstantially dissolving the remaining target in the source material.For example, in one embodiment the source is in a solid phase andremains in the solid phase throughout the irradiation and separationoperations.

In the embodiment illustrated in FIG. 12, an optional re-irradiationoperation 1210 may be performed. In that operation, the sources areexposed to neutrons for some irradiation period in an irradiationoperation 1210. This operation 1210 may include transporting thesource(s) to the irradiation facility/equipment for safe irradiation,for example by conveyor belt as described above. In the irradiationoperation 1210, source material is exposed to neutrons, thereby causingat least some atoms of the source material to undergo nuclear fission orneutron capture to create atoms of the selected radioisotope. Thisresults in a re-irradiated source material that contains some amount ofthe selected radioisotope product within a reduced amount of unreactedtarget as discussed with reference to FIG. 11. In addition, because ofthe recoil from the fission reaction, at least some of the newly createdatoms of the selected radioisotope move the recoil distance relative tothe remaining, unreacted target within the source material. As describedabove, the recoil of the selected radioisotope product may make thatradioisotope more available to the extraction material such as by makingthe radioisotope product closer to an available surface of the sourcematerial, which may then improve extraction by the extraction material.

In embodiments in which sources include a container, the source materialmay or may not be removed from the container during the separationoperation 1208. The separation operation 1208 may further includeregeneration of the target to prepare it for subsequent irradiation.This may involve one or more washing operations to remove extractionmaterial from the source material prior to subsequent irradiation.

After a separation operation 1208, the same source may be re-irradiatedto create more of the selected radioisotope allowing the irradiation andseparation operations 1210, 1208 to be repeated multiple times withoutsubstantially dissolving, changing or removing any of the mass ofremaining target material in the source (except as a result of thefusion reaction). As discussed above, this allows the fissionablematerial to be more efficiently converted into the desired product thanwould be possible with a single neutron exposure. The method 1200further includes a final processing operation 1214 that converts theextracted radioisotope product into a final product or final formsuitable for commercial use, as described with reference to FIG. 3. Inan embodiment, the final processing operation 1214 includesincorporating the radioisotope into a daughter isotope generator asdescribed above.

It will be clear that the systems and methods described herein are welladapted to attain the ends and advantages mentioned as well as thoseinherent therein. Those skilled in the art will recognize that themethods and systems within this specification may be implemented in manymanners and as such is not to be limited by the foregoing exemplifiedembodiments and examples. In this regard, any number of the features ofthe different embodiments described herein may be combined into onesingle embodiment and alternate embodiments having fewer than or morethan all of the features herein described are possible.

While various embodiments have been described for purposes of thisdisclosure, various changes and modifications may be made which are wellwithin the scope of the technology described herein. For example,targetry coupled separation may be adapted to remove fission productsincluding poisons or other nuclear contaminants from sources made ofsolid nuclear waste. Numerous other changes may be made which willreadily suggest themselves to those skilled in the art and which areencompassed in the spirit of the disclosure and as defined in theappended claims.

1-20. (canceled)
 21. A method of manufacturing a radionuclide metal thatis a fission product of the fissioning of uranium, the methodcomprising: fissioning a molten fuel salt containing uranium, therebygenerating an irradiated fuel salt mixture containing molten uraniumsalt and fission products, wherein the fission products include theradionuclide metal; contacting at least some of the irradiated fuel saltmixture with supercritical carbon dioxide containing a ligand that formsa metal complex with the radionuclide metal but does not form a metalcomplex with the uranium in the fuel salt, thereby forming a combinedfuel salt and supercritical carbon dioxide mixture containing an amountof radionuclide metal complexes; separating at least some of theradionuclide metal complexes from the combined fuel salt andsupercritical carbon dioxide mixture; and decomposing the radionuclidemetal complexes to obtain the radionuclide metal.
 22. The method ofclaim 21, wherein the contacting further comprises: removing theirradiated fuel salt mixture from a molten fuel salt nuclear reactor;placing the irradiated fuel salt mixture in an extraction vessel; andinjecting the supercritical carbon dioxide containing the ligand intothe extraction vessel.
 23. The method of claim 22, wherein thecontacting further comprises: placing the irradiated fuel salt mixtureinto the extraction vessel containing supercritical carbon dioxidecontaining the ligand.
 24. The method of claim 21, wherein theseparating further comprises: adjusting at least one of a temperature ora pressure of the combined fuel salt and supercritical carbon dioxidemixture in an extraction vessel.
 25. The method of claim 24, wherein theadjusting further comprises: increasing the temperature, the pressure,or both to volatilize the radionuclide metal complexes out of thecombined fuel salt and supercritical carbon dioxide mixture as aseparate gas phase.
 26. The method of claim 25, wherein the separatingfurther comprises: removing the separate gas phase from the extractionvessel.
 27. The method of claim 26, wherein the decomposing furthercomprises: heating the separate gas phase, thereby decomposing theradionuclide metal complexes.
 28. The method of claim 24, wherein theadjusting further comprises: decreasing the temperature, the pressure,or both to obtain a liquid phase separate from the combined fuel saltand supercritical carbon dioxide mixture, the liquid phase containingthe radionuclide metal complexes.
 29. The method of claim 28, whereinthe separating further comprises: removing the liquid phase from theextraction vessel.
 30. The method of claim 29, wherein the decomposingfurther comprises: heating the liquid phase, thereby decomposing theradionuclide metal complexes.
 31. The method of claim 21, wherein theradionuclide metal is selected from ²²⁷Ac, ²¹³Bi, ¹³¹Cs, ¹³³Cs, ¹¹C,⁵¹Cr, ⁵⁷Co, ⁶⁰Co, ⁶⁴Cu, ⁶⁷Cu, ¹⁶⁵Dy, ¹⁶⁹Er, ¹⁸F, ⁶⁷Ga, ⁶⁸Ga, ⁶⁸Ge,¹⁹⁸Au, ¹⁶⁶Ho, ¹¹¹In, ¹²³I, ¹²⁴I, ¹²⁵I, ¹³¹I, ¹⁹²Ir, ⁵⁹Fe, ⁸¹mKr, ²¹²Pb,¹⁷⁷Lu, ⁹⁹Mo, ¹³N, ¹⁵O, ¹⁰³Pd, ³²P, ²³⁸Pu, ⁴²K, ²²⁷Ra, ²²³Ra, ¹⁸⁶Re,¹⁸⁸Re, ⁸¹Rb, ⁸²Rb, ¹⁰¹Ru, ¹⁰³Ru, ¹⁵³Sm, ⁷⁵Se, ²⁴Na, ⁸²Sr, ⁸⁹Sr, ⁹⁹mTc,and ²⁰¹Tl.
 32. The method of claim 21, wherein the molten fuel saltcontains a chloride salt of uranium.
 33. The method of claim 21, whereinthe contacting is performed with the irradiated fuel salt mixture in aliquid form.
 34. The method of claim 21, wherein the contacting isperformed with the irradiated fuel salt mixture in a solid form.
 35. Themethod of claim 21, wherein the contacting is performed on theirradiated fuel salt mixture while the irradiated fuel salt mixture isin a molten fuel salt nuclear reactor.
 36. A method of manufacturing¹⁷⁷Lu radioisotope comprised of: preparing an enriched dry ¹⁷⁶Yb oxidesource material within a pressurized containment vessel; irradiating the¹⁷⁶Yb oxide containing source with neutrons at suitable neutron fluxdensity to achieve reaction of ¹⁷⁶Yb to ¹⁷⁷Yb via the reaction¹⁷⁶Yb(n,γ)¹⁷⁷Yb from which ¹⁷⁷Yb will decay to ¹⁷⁷Lu by the reaction¹⁷⁷Yb→¹⁷⁷Lu+β⁻ with half-life of 1.91 h; injecting the pressurizedcontainment vessel containing the irradiated ¹⁷⁶Yb oxide source withsupercritical carbon dioxide containing a ligand that selectively formsa metal complex with the lutetium but does not form a metal complex withthe ytterbium, thereby forming a supercritical carbon dioxide mixturecontaining an amount of lutetium metal complexes; removing thesupercritical carbon dioxide mixture containing an amount of lutetiummetal complexes; and separating at least some of the radionuclidelutetium complexes from the supercritical carbon dioxide mixture toprovide a ¹⁷⁷Lu rich separation fraction.
 37. A method of manufacturing⁶⁷Cu radioisotope comprised of: packing an enriched ⁷⁰Zn oxide sourcematerial in an appropriate a cavity within an irradiation target box;irradiating the ⁷⁰Zn source with 15-20 MeV of deuterons), yielding ⁶⁷Cuvia the reaction ⁷⁰Zn(p,α)⁶⁷Cu with estimated cross-sections of ⁶⁷Cuformation of 15-27 mb; contacting the ⁷⁰Zn oxide target withsupercritical carbon dioxide containing a ligand that selectively formsa metal complex with the copper but does not form a metal complex withthe zinc oxide, thereby forming a supercritical carbon dioxide mixturecontaining an amount of radionuclide metal complexes; separating atleast some of the radionuclide metal complexes from the combinedradionuclide and supercritical carbon dioxide mixture; separating theradionuclide from the combined radionuclide and supercritical carbondioxide mixture; and purifying the ⁶⁷Cu radionuclide.